ML17150A088
| ML17150A088 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/25/2017 |
| From: | Bielby M Operations Branch III |
| To: | NRC/RGN-III/DRS/OLB |
| Shared Package | |
| ML17061A756 | List: |
| References | |
| Download: ML17150A088 (207) | |
Text
1 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The following sequence of events occurs in the plant:
Time Event 1000 The reactor is operating at 100% power.
Total Core Flow is 85%.
1005 RR MG A speed is 70% and unexpectedly LOWERING.
1010 3D108, RECIRC SYS A FLUID DRIVE LUBE OIL PRESS LOW, alarms.
1015 The reactor is operating at 80% power.
Total Core Flow is 66%.
1020 3D104, RECIRC PUMP A MG DRIVE MTR GRD FAULT / OVERCURRENT, alarms.
Based on the above conditions, 3D112, RECIRC SYS A FLUID DRIVE SCOOP TUBE LOCK, alarmed at (1) .
At 1020, Total Core Flow INDICATION (2) .
A. (1) 1005 (2) INCREASED due to lower Loop A Flow Resistance B. (1) 1005 (2) DECREASED due to subtraction of Loop A Flow from Loop B C. (1) 1010 (2) INCREASED due to lower Loop A Flow Resistance D. (1) 1010 (2) DECREASED due to subtraction of Loop A Flow from Loop B Answer: B 2017 ILO Written Page: 1 of 207 Question 1
Answer Explanation:
B. Criteria for Automatic Scoop Tube Lockup on control deviation (5%), and Loop Flow Subtraction occurs when RR Pump Trips.
1010 Low Oil Pressure will start RR MG Emergency Oil Pump C, and would cause a Scoop Tube Lock but it has already occurred. Reductions in Recirc flow resistance normally raises actual core flow during a power reduction by rods, the flow subtraction network causes INDICATED Core Flow to DECREASE.
Plant Procedures 03D112 NUREG 1123 KA Catalog Rev. 2 295001 AK2. Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following:
295001 AK2.01 Recirculation system 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
NRC Exam Usage ILO 2017 Exam LOR 2009 Exam LOR 2015 Exam ILO 2017 NRC Question Use:
Bank High RO 2017 ILO Written Page: 2 of 207 Question 1
2 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: BANK The plant is operating at 90% power when an event occurs and the following indications are seen in the control room:
- 120 kV BUS 101 POWER ON light OFF
- 120 kV BUS 102 POWER ON light OFF
- R14-R819, SS Xfmr # 64 Secondary Volt Indicator 0 (zero) volts
- 345 kV BUS 301 POWER ON light OFF
- 345 kV BUS 302 POWER ON light OFF
- R14-R833, SS Xfmr # 65 Secondary Volt Indicator 0 (zero) volts
- R14-R835, SS Xfmr # 65 Secondary Volt Indicator 0 (zero) volts Given these conditions, which one of the following actions is required, and why?
A. Place the Reactor Mode Switch in SHUTDOWN due to a COMPLETE loss of Offsite Power.
B. Manually start and load ALL Emergency Diesel Generators due to loss of Offsite Power to BOTH divisions of ESF Buses.
C. Place the Reactor Mode Switch in SHUTDOWN due to loss of Offsite Power to ONLY ONE division of ESF Buses AND half of the BOP Buses.
D. Verify automatic start and loading of Emergency Diesel Generators 11 AND 12 due to loss of Offsite Power source to ONLY ONE division of ESF Buses.
Answer: A 2017 ILO Written Page: 3 of 207 Question 2
Answer Explanation:
Stem conditions indicate complete Loss of Offsite Power. 20.300.Offsite Immediate Action requires placing Reactor Mode Switch in SHUTDOWN.
Distractor Explanation:
B. Is plausible and incorrect, EDGs should have auto started and it is required to verify automatic EDG starts. It is not appropriate to manually start EDGs.
C. Is plausible and incorrect, if power was lost to only the 345kV system, ONE Division of ESF Buses and Half of the BOP Buses would be lost. Mode switch to shutdown is required.
D. Is plausible and incorrect, if power were lost to only the 120 kV system, ONE Division of ESF Buses would be affected.
Reference Information:
20.300.Offsite IA (pg_3)
Plant Procedures 04D132 20.300.Offsite NUREG 1123 KA Catalog Rev. 2 295003 AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER :
295003 AK3.05 3.7/3.7 Reactor SCRAM 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
NRC Exam Usage ILO 2017 Exam LOR 2009 Exam ILO 2017 NRC Question Use:
Bank High RO 2017 ILO Written Page: 4 of 207 Question 2
3 Points: 1.00 Difficulty: 4.00 Level of Knowledge: High Source: BANK The plant is operating at 100% power with RHR Pump "A" operating in Torus Cooling. A transient occurs, resulting in the following conditions:
- Control Rods All fully inserted
- MSIVs All closed
- RCIC System injecting at 650 gpm
- 1D6, DIV I CSS LOGIC POWER FAILURE Alarming
- 9D17, DIV I ESS 130V BATTERY TROUBLE Alarming
- ALL Div. 1 ECCS Pumps CMC indication lost (no lights)
Based on these conditions, RHR Pump "A" is (1) and (2) .
A. (1) RUNNING (2) CAN be electrically tripped B. (1) RUNNING (2) CANNOT be electrically tripped C. (1) TRIPPED (2) WILL auto start on receipt of a valid signal D. (1) TRIPPED (2) WILL NOT auto start on receipt of a valid signal Answer: B 2017 ILO Written Page: 5 of 207 Question 3
Answer Explanation:
B. RHR Pump "A" has lost control power, indicated by the DC Annunciators; therefore RHR Pump "A" will continue to run and cannot be electrically tripped. Control power is required for both CMC indication and trip function.
Distractor Explanation:
A. Is plausible and incorrect; RHR pump will remain running, but cannot be electrically tripped.
C. Is plausible and incorrect; RHR pump loses control power, which is required for BOTH trip and start functions.
D. Is plausible and incorrect; RHR pump loses control power, which is required for BOTH trip and start functions.
Reference Information:
ARP 1D8 (pg 1)
Plant Procedures 01D08 NUREG 1123 KA Catalog Rev. 2 295004 AA1. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER :
295004 AA1.02 Systems necessary to assure safe plant shutdown 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
NRC Exam Usage ILO 2017 Exam LOR 2009 Exam LOR 2015 Exam I-Drawings I-2201-01 ILO 2017 NRC Question Use:
Bank High RO 2017 ILO Written Page: 6 of 207 Question 3
4 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is operating at 90% power. 24.110.05, RPS - Turbine Control and Stop Valve Functional Test is in progress. One portion of the test did not result in a Half Scram indication when required.
The operator action which results in the HIGHEST power condition that meets LCO requirements is to lower reactor power to (1) based on a turbine trip violating the (2) safety limit.
A. (1) 26%
(2) Minimum Critical Power Ratio B. (1) 29%
(2) Minimum Critical Power Ratio C. (1) 26%
(2) Reactor Coolant System Pressure D. (1) 29%
(2) Reactor Coolant System Pressure Answer: B 2017 ILO Written Page: 7 of 207 Question 4
Answer Explanation:
Per the DBD (C71-00) An automatic bypass shall be provided for both the turbine stop valve closure and turbine control valve fast closure trips when the turbine first stage pressure is equal to, or less than a value that is equivalent to 30 percent nuclear boiler rated power. The DBD further states that the RPS Trip Function of TSV Closure is based on operational and safety limits (MCPR)
Further explanation can be found in TS & TSB Below 29.5%, the RPS Trip Function of TSV Closure is Not Applicable. This is a required ACTION of T.S.
3.3.1 D.1.E.
TSV Closure Trip is based on not exceeding Minimum Critical Power Ratio Safety Limit. (TSB 3.3.1)
Distractor Explanation:
Reactor Coolant System Pressure Safety Limit is considered plausible since a Turbine Trip without Bypass ATWS may cause a rapid reduction in steam flow accompanied by power rise from void collapse.
Reference Information:
C71-00 DBD REACTOR PROTECTION SYSTEM Objective Link LP-OP-315-0129-C004. State major precautions and limitations, and major safety considerations for the Turbine Steam System, and describe their bases.
Plant Procedures 20.109.01 NUREG 1123 KA Catalog Rev. 2 241000 K4. Knowledge of REACTOR/TURBINE PRESSURE REGULATING SYSTEM design feature(s) and/or interlocks which provide for the following:
241000 K4.05 Reactor scram 295005 AA2. Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP :
295005 AA2.05 Reactor power 295005 AK2. Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following:
295005 AK2.01 RPS Technical Specifications 3.3.1.1 Reactor Protection System (RPS) Instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
NRC Exam Usage ILO 2017 Exam LOR 2009 Exam LOR 2015 Exam ILO 2017 NRC Question Use:
Bank High RO 2017 ILO Written Page: 8 of 207 Question 4
5 Points: 1.00 Difficulty: 3.50 Level of Knowledge: High Source: NEW The plant was operating at 20% power during a startup with Main Turbine warmup in progress. A pressure regulator failure HIGH occurs resulting in a Reactor Scram. NO operator actions have been taken.
Given the above conditions, manual operation of ____ could remove Decay Heat, without any other system manipulation.
B. RCIC for level control.
C. Main Steam Line Drain Valves.
D. Moisture Separator Reheater Isolation valves.
Answer: B 2017 ILO Written Page: 9 of 207 Question 5
Answer Explanation:
With power at 20%, the Mode Switch is in RUN. Pressure regulator failure high causes the Main Turbine Bypass Valves to fully open, lowering RPV Pressure. When pressure reaches 756 psig with Mode Switch in RUN, a GP1 PCIS isolation occurs and a Scram. Decay heat from these conditions will not result in reactor pressure rising to 1093 psig for Low-Low-Set. With MSIVs closed, only RWCU is immediately available to remove decay heat.
Distractor Explanation:
A. and C. Are incorrect and plausible since the Turbine BPVs and MSL Drains are not available due to the Gp1 Isolation.
D. Is incorrect plausible because of the Gp1 isolation and opening the MSR valves could cause a sudden Turbine speed increase.
Reference Information:
20.000.21 Bases (general info on scram) 23.601 Low pressure Isolations (pg 9) 23.601 SCRAM cause (pg 28)
Plant Procedures 20.000.21 20.000.21 Bases NUREG 1123 KA Catalog Rev. 2 295006 AK1. Knowledge of the operational implications of the following concepts as they apply to SCRAM :
295006 AK1.01 Decay heat generation and removal 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
NRC Exam Usage ILO 2017 Exam ILO 2017 NRC Question Use:
High New RO 2017 ILO Written Page: 10 of 207 Question 5
6 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: BANK Which one of the following correctly identifies the RCIC controls or indications available BOTH in the Main Control Room AND at the Remote Shutdown Panel?
A. RCIC Pump Flow indication.
B. RCIC Pump Discharge Pressure indication.
C. E5150-F010, RCIC Pump CST Suction Isolation Valve, control pushbuttons.
D. RCIC Pump Flow controller with Manual and Automatic Setpoint Adjustments.
Answer: A 2017 ILO Written Page: 11 of 207 Question 6
Answer Explanation:
RCIC Flow Indication exists at both locations.
Distractor Explanation:
B. Is incorrect and plausible because it exists only in the MCR.
C. Is incorrect and plausible because it exists only in the MCR.
D. Is incorrect and plausible because it exists only in the MCR.
Reference Information:
20.000.19, page 6 Plant Procedures 20.000.19 NUREG 1123 KA Catalog Rev. 2 295016 AK2. Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:
295016 AK2.01 Remote shutdown panel: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.
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Bank Low RO Selected for Early Review (NRC) 2017 ILO Written Page: 12 of 207 Question 6
7 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant was operating at 65% when an event results in the following conditions:
- North TBCCW pump Tripped
- Center TBCCW pump Tripped
- 480V Bus 72N Deenergized
- Stator Water Cooling Generator Inlet Temperature (Field Report) 130°F rising
- TBCCW Head Tank Level (Field Report) Empty
- N & S RFP & RFPT Turbine Bearing Drain Temperatures 160 to 165°F rising The CRS directs placing the Mode Switch in Shutdown and the trip the Main Turbine. What is the basis for the directed actions?
A. The only basis for this action is to place the reactor in a safe condition and prevents damage to Main Turbine Generator equipment.
B. The only basis for this action is to preventing overheat of the Reactor Feed Pump bearings.
C. The only basis for this action is to prevent damage to the Main Turbine Generator.
D. The only basis for this action is to place the reactor in a safe condition.
Answer: A 2017 ILO Written Page: 13 of 207 Question 7
Answer Explanation:
The above conditions require entry into 20.128.01, Loss of Turbine Building Cooling Water. Since the N &
C TBCCW pumps are tripped and the S TBCCW pump is deenergized (loss of 72N), a total loss of TBCCW has occurred. The first AOP override requires placing the Mode Switch in S/D and tripping the MTG. These actions insert a reactor scram to bypass MSIV isolation on low pressure, and trip the MTG to prevent exceeding design temperatures, preventing equipment damage.
Distractor Explanation:
Preventing overheat of the Reactor Feed Pump bearings, preventing damage to the Main Turbine Generator, and placing the reactor in a safe condition are all items that the overides are designed for. The CRS's use of the overides is for all, not just one.
Reference Information:
20.128.01 Bases OVERRIDE (Pg 2) 20.128.01 SYMPTOMS (Pg 9)
Plant Procedures 20.128.01 20.128.01 Bases NUREG 1123 KA Catalog Rev. 2 295018 AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:
295018 AK3.02 Reactor power reduction 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.
10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
NRC Exam Usage ILO 2017 Exam ILO 2017 NRC Question Use:
High New RO 2017 ILO Written Page: 14 of 207 Question 7
8 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant is operating normally at 100% power when a system air rupture occurs. It takes the CRLNO several minutes to get to the P807 panel. No operator actions associated with the air systems have been performed.
(Review HANDOUT: Air System Mimic from H11-P807)
Based on only these indications, where is the location of the air rupture?
A. IAS Header B. D1 NIAS Header C. D2 NIAS Header D. Turbine Bldg Header Answer: D 2017 ILO Written Page: 15 of 207 Question 8
Answer Explanation:
The air rupture location can be identified using air valve position indication on the P807 panel. At 100%
power, the only valves closed are P5000-F403, IAS to Div 2 NIAS, and the DW SA Supply isolation valves, P5000-F603 and F604. A rupture in the Turbine Bldg air header will result in lowering Station Air Header pressure. At <95 psig, the standby air compressor will auto-start, and at <85 psig P5000-F401, Station Air to Turbine Bldg Header isolation valve, will close which isolates the air leak allowing Station Air Header pressure to recover. All other open isolation valves remain open.
Distractor Explanation:
A. Is plausible if the assumption is made that Station Air pressure continues to lower to <75 psig which would result in the remaining isolation valves to close.
B. Is plausible if the assumption is made that P5000-F403, IAS to Div 2 NIAS isolation valve, has closed due to a leak in Div 2 NIAS.
C. Is plausible if the assumption is made that P5000-F401, Station Air to TB Hdr isolation valve, will isolate a leak in the Station Air Header.
References Information:
I-2450-04, 05 Plant Procedures 20.129.01 23.129 NUREG 1123 KA Catalog Rev. 2 295019 Partial or Complete Loss of Instrument Air G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
NRC Exam Usage ILO 2017 Exam I-Drawings I-2450-04 I-2450-05 ILO 2017 NRC Question Use:
High New RO 2017 ILO Written Page: 16 of 207 Question 8
9 Points: 1.00 Difficulty: 4.00 Level of Knowledge: High Source: NEW The plant is in MODE 5 with fuel handling operations in progress and Division 1 RHR operating in Shutdown Cooling.
The operating RHR Pump trips when an electrical fault causes the closure of E1150-F008, RHR SDC Otbd Suction Isol Valve. The valve cannot be reopened electrically or manually.
Which ONE of the following actions is required under these conditions?
A. Place Division 2 RHR in Shutdown Cooling.
B. Start ONE RHR Pump and open SRVs to drain water to the Torus.
C. Shift Fuel Pool Cooling and Cleanup discharge to the Reactor Well.
D. Start ONE Core Spray Pump and open SRVs to drain water to the Torus.
Answer: C 2017 ILO Written Page: 17 of 207 Question 9
Answer Explanation:
BOTH Loops of RHR are rendered unavailable by the COMMON suction line isolation. With NO RHR available and the Reactor Vessel Head and Fuel Pool Gates removed, 20.205.01, Loss of Shutdown Cooling, directs shifting FPCCU discharge to the Reactor Well, maximizing flow to the Reactor Cavity.
Distractors Explanation:
B & D are plausible and incorrect; If the Head and Fuel Pool Gates were in place and MS line plugs removed these would be correct actions.
A is plausible and incorrect; Since F008 is on the COMMON suction line, with no bypass valve, with no bypass valve DIV 2 RHR will not be available.
Reference Information:
20.205.01 Condition J (pg 9)
Plant Procedures 20.205.01 NUREG 1123 KA Catalog Rev. 2 295021 AK2. Knowledge of the interrelations between LOSS OF SHUTDOWN COOLING and the following:
295021 AK2.05 Fuel pool cooling and cleanup system 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High New RO 2017 ILO Written Page: 18 of 207 Question 9
10 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW An irradiated fuel bundle is being removed from the core. An adjacent bundle has been lifted along with the selected bundle. When noticed, fuel movement is stopped, and the adjacent bundle falls back into the core.
Bubbles come to the pool surface, and the local Continuous Air Monitor (CAM) alarms. The following indications are observed:
- 16D1, RB REFUELING AREA FIFTH FLOOR HIGH RADN, alarms.
- D21-R600 Radiation Monitoring Recorder, Channel 17 indicates 10 mr/hr.
- D21-K717, RB5 Refuel Floor Lo Range ARM Ind trip Unit (Ch. 17), indicates tripped.
Which of the following is (1) an automatic action (if any) that will occur as a result of this event and (2) the reason for the action?
A. (1) Radiation alarm ONLY (2) Warns personnel of possible hazards and to evacuate affected area B. (1) ONE division of RBHVAC trips, and ONE division of SGTS Auto Starts (2) Ensures negative pressure maintained in Secondary Containment, and prevents exceeding system design flow rate C. (1) BOTH divisions of RBHVAC trip, and BOTH divisions of SGTS Auto Start (2) Stops the unfiltered release of radioactive particles to the environment, and maintains negative pressure in the Reactor Building D. (1) Group 14-Drywell and Suppression Pool Ventilation System, and Group 16-Nitrogen Inerting System isolations ONLY (2) Closes off release routes for radioactive materials from the Secondary Containment to contain and limit the release to the environment Answer: A 2017 ILO Written Page: 19 of 207 Question 10
Answer Explanation:
Area radiation monitors provide alarm only, and FP Vent Rad Monitor upscale provides no automatic actions. 16D1 and 3D31 require AOP entry and plant announcement for personnel evacuation. Also, the indications provided DO NOT show high level trips of the required instrumentation (3D35/3D36) for RBHVAC trip/SGTS actuation.
Distractor Explanation:
B. Is incorrect but plausible if the candidate assumes the indications show a high level trip of one trip channel of the Fuel Pool Vent Exhaust Rad Monitor (only one division actuated).
C. Is incorrect but plausible if the candidate assumes the indications show a high level trip of at least two trip channels of the Fuel Pool Vent Exhaust Rad Monitor (both divisions actuated).
D. Is incorrect but plausible because these actions are associated with Reactor Building Vent Exhaust Rad Monitor Upscale Trip or Fuel Pool Vent Exhaust Rad Monitor Upscale Trip, but are not the only actions for the trip (if actuated).
Reference Information:
20.710.01 (pg 3)
Plant Procedures 20.710.01 20.710.01 Bases NUREG 1123 KA Catalog Rev. 2 295023 AK3. Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS :
295023 AK3.03 Ventilation isolation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(11) Purpose and operation of radiation monitoring systems, including alarms and survey equipment.
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Low New RO 2017 ILO Written Page: 20 of 207 Question 10
11 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is being started after a refueling outage. Reactor power is 9% when a small steam leak develops in the drywell. The PCRMS is in service with no alarms indicated.
The CRS enters 29.100.01, Sheet 2, Primary Containment Control, based on high Drywell temperature of 145°F. Drywell pressure is 0.7 psig and rising 0.05 psig every 15 minutes.
Based on these plant conditions what action(s) should be taken?
A. Start SGTS, and vent the Torus.
B. Place RHR in Torus Cooling and Torus Spray.
C. Verify RBHVAC is running, and vent the Drywell.
D. Scram the reactor in anticipation of an automatic scram.
Answer: C 2017 ILO Written Page: 21 of 207 Question 11
Answer Explanation:
The EOPs, step PCP-1, direct venting the drywell to keep DW pressure <1.68 psig in accordance with 29.ESP.07, step 1.0. The candidate should recognize that DW pressure is high but not rising fast, so venting the DW is appropriate.
Distractor Explanation:
A. Is plausible and incorrect because a steam leak would indicate particulate in the air. The candidate could expect the vent path to be through SGTS to reduce radioactive release, however, there are no alarms on PCRMS which indicates no (or little) radioactive particulates to scrub.
B. Is plausible and incorrect because there are elevated containment parameters. The candidate may determine that containment cooling for pressure reduction is necessary at this time.
D. Is plausible and incorrect because the EOPs have been entered. The candidate may deduce that a scram on high DW pressure is imminent and that scramming the reactor is a conservative action. This is not the correct answer because venting the drywell is directed and should allow the operators to perform a controlled shutdown without placing the plant through a scram transient.
Reference Information:
29.ESP.07 (pg 3)
Plant Procedures 29.ESP.07 NUREG 1123 KA Catalog Rev. 2 295024 EA1. Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE:
295024 EA1.14 Drywell ventilation system 295024 EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:
295024 EA2.01 Drywell pressure.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank High RO 2017 ILO Written Page: 22 of 207 Question 11
12 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant is operating at 80% power with the standby Reactor Pressure Regulator out of service for repairs. The in-service Reactor Pressure Regulator then malfunctions with its output slowly lowering.
Assuming no operator intervention, reactor power will ___________________________.
A. lower until the MSIVs isolate on low steam pressure.
B. remain constant as the Turbine Bypass Valves open to control reactor pressure.
C. rise, MWe output will lower, and the reactor will scram on high pressure OR high power.
D. stabilize at a lower level due to a Reactor Recirculation Pump runback from the loss of FW heating.
Answer: C 2017 ILO Written Page: 23 of 207 Question 12
Answer Explanation:
Pressure Regulator failure with lowering output results in reactor pressure rising. Turbine Bypass Valves remain closed and Turbine Control Valves close, causing Turbine Generator output (MWe) to lower.
Reactor pressure will rise until the reactor scrams on high reactor pressure or high reactor power.
Distractor Explanation:
A. Is plausible if the candidate thinks lowering pressure regulator output results in lowering reactor pressure. This is incorrect since the failure causes reactor pressure to rise.
B. Is plausible if the candidate thinks the Turbine Bypass Valves will operate to lower reactor pressure.
This is incorrect since the failed regulator will maintain the bypass valves closed.
D. Is plausible if the candidate thinks the pressure change will result in a loss of FW heating as a result of the sensitivity of the heater drains system and Recirc Runback. This is incorrect because the reactor will scram due to the rising pressure.
Reference Information:
20.109.02 Enclosure A 23.601 (pg 10 & pg 41)
Plant Procedures 20.109.02 23.109 NUREG 1123 KA Catalog Rev. 2 262002 K1. Knowledge of the physical connections and/or cause-effect relationships between UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) and the following:
262002 K1.03 Rod position information: Plant-Specific 295025 EK1. Knowledge of the operational implications of the following concepts as they apply to High Reactor Pressure:
295025 EK1.01 Pressure effects on reactor power 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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High New RO 2017 ILO Written Page: 24 of 207 Question 12
13 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED After a plant transient and reactor scram, the following conditions exist:
- Div 1 RHR is being used for Torus Cooling/Torus Spray with ONE Pump at 11,500 gpm.
- Div 1 Core Spray is injecting with TWO Pumps at 7,750 gpm.
- Torus Pressure is 5.0 psig.
- Torus Level is -70 (minus 70) inches.
- Torus Temperature is 190°F.
The NO in the Reactor Building calls to report the RHR and Core Spray Pumps are rattling.
Using current pump configuration, which of the following is the highest flow permissible?
A. Core Spray Flow - 5,500 gpm; RHR Flow - 8,000 gpm B. Core Spray Flow - 6,500 gpm; RHR Flow - 10,000 gpm C. Core Spray Flow - 6,500 gpm; RHR Flow - 8,000 gpm D. Core Spray Flow - 5,500 gpm; RHR Flow - 10,000 gpm Answer: C 2017 ILO Written Page: 25 of 207 Question 13
Answer Explanation:
Candidate must apply NPSH and Vortex limits. Torus Overpressure calculates to be (5+3.5+((-
70)/30))=6.2 psig. This means that the 5 PSIG overpressure curve applies.
The RHR vortex limit for -70 is <9,000 gpm, while NPSH is 12,000 gpm.
The CS vortex limit for -70 is <7,000 gpm, while NPSH is 8150 gpm (Max).
The CS flow is limited by Vortex limit curve (<7000), and RHR flow is limited by Vortex limit curve(<9000).
Distractor Explanation:
All distractors are plausible because they contain a violation of one or both curves.
Reference Information:
29.100.01 SH 6 2017 ILO Written Page: 26 of 207 Question 13
Plant Procedures 29.100.01 SH 6 NUREG 1123 KA Catalog Rev. 2 295026 EK1. Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE :
295026 EK1.01 Pump NPSH 295026 EK2. Knowledge of the interrelations between SUPPRESSION POOL HIGH WATER TEMPERATURE and the following:
295026 EK2.02 Suppression pool spray: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.
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High Modified RO 2017 ILO Written Page: 27 of 207 Question 13
14 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: MODIFIED At which of the following Drywell Temperature and RPV pressure combinations may RPV Water level indication become unreliable due to boiling in the instrument run?
Drywell Temperature RPV pressure A. 250°F 25 PSIG B. 300°F 75 PSIG C. 350°F 100 PSIG D. 375°F 250 PSIG Answer: C 2017 ILO Written Page: 28 of 207 Question 14
Answer Explanation:
Caution 1 of 29.100.01, Sheet 6, states that RPV level indications are affected by instrument run temperatures and RPV pressure, and that the instrument may become unreliable due to boiling in the run if the temperature near the run is above the RPV saturation temperature. The candidate must use the RPV Saturation Temperature curve to determine which of the given sets of temperatures/pressures put conditions in the unreliable water level area of the curve. According to the curve, RPV level indication may become unreliable if RPV pressure is 100 psig and Drywell Temperature is greater than ~330°F. This makes distracter C correct since pressure is 100 psig and temperature is 350°F.
Distractor Explanation:
The other distractors are incorrect but plausible based in plotting the data in the RPV Saturation Temperature curve.
Reference Information:
29.100.01 SH 6 Plant Procedures 29.100.01 SH 6 NUREG 1123 KA Catalog Rev. 2 295028 EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE :
295028 EA2.03 Reactor water level 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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High Modified RO 2017 ILO Written Page: 29 of 207 Question 14
15 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The following plant conditions exist after a LOCA event:
- Torus Water Temperature 180°F (stable)
- Torus Water Level -30 inches (stable)
- Torus Pressure 0 psig (stable)
- Reactor Pressure 70 psig (stable)
- CST Level 0.1 feet Which one of the following describes ACCEPTABLE conditions for the following pumps being used for RPV Level control?
A. RCIC operation at 650 gpm.
B. HPCI operation at 4500 gpm.
C. Single pump LPCI Flow of 12,500 gpm.
D. Total Core Spray Loop Flow of 6800 gpm.
Answer: D 2017 ILO Written Page: 30 of 207 Question 15
Answer Explanation:
The maximum Core Spray flow based on Torus temperature and pressure is found using the Core Spray NPSH Limit curve. At 180°F Torus Temperature and 2.5 psig Torus pressure (torus overpressure), loop flow is limited to <7000 gpm.
Distractor Explanation:
A. Is plausible and incorrect, because RCIC Flow Limit is <600 gpm at 180°F Torus Temperature.
B. Is plausible and incorrect, because the HPCI Turbine is isolated on Low Steam Press, and flow is limited to <3500 gpm at 180°F Torus Temperature.
C. Is plausible and incorrect, because RHR Loop Flow of <10,000 gpm is the NPSH Limit at 180°F Torus Temperature.
Reference Information:
29.100.01 SH 6 ILO Exam Note:
Double Jeopardy with R13 discussion. For this question Distractors 1 & 2 (A&B) are not on R13. For Distractor 3 (C) and the answer, R13 required correctly reading Vortex Limit. This question requires correctly reading NPSH. While NPSH is reviewed on R13, it is NOT the limiting factor.
Plant Procedures 29.100.01 SH 6 NUREG 1123 KA Catalog Rev. 2 295030 Low Suppression Pool Water Level 295030 EK1. Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL:
295030 EK1.02 Pump NPSH 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank High RO 2017 ILO Written Page: 31 of 207 Question 15
16 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant is operating at 100% power with Feedwater Control System in Auto (3 element) when RPV level falls below 173". All actuations occur as expected.
What is the status of the Feed Water Level Control System?
A. C32-R618, Master Feedwater Level Controller, setpoint setdown to 150 inches and C32-R620, N21-F403 Startup LCV Controller at 197 in single element control.
B. C32-R618, Master Feedwater Level Controller, setpoint setdown to 197 inches and C32-R620, N21-F403 Startup LCV Controller at 197 in three element control.
C. C32-K616A, North Reactor Feed Pump Controller, and C32-K616B, South Reactor Feed Pump Controller, in AUTO with outputs corresponding to approximately 3200 rpm.
D. C32-R618, Master Feedwater Level Controller, setpoint at 197 inches with C32-K616A, North Reactor Feed Pump Controller, and C32-K616B, South Reactor Feed Pump Controller AUTO.
Answer: A 2017 ILO Written Page: 32 of 207 Question 16
Answer Explanation:
Post-Scram Feedwater Logic initiation from a Reactor Scram caused by low RPV level.
Distractor Explanation:
B. Is incorrect and plausible because setpoint setdown is 150 inches and stays in three element for a Reactor Scram caused by low reactor water level.
C. Is incorrect and plausible because Feed pump controllers are pulsed to manual with an output of approximately 2650.
D. Is incorrect and plausible because Normal reactor water level control indications with no Reactor Scram.
Reference Information:
ARP 3D157 (pg 1)
Plant Procedures 03D157 NUREG 1123 KA Catalog Rev. 2 295031 EK2. Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following:
295031 EK2.16 Reactor water level control 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High New RO 2017 ILO Written Page: 33 of 207 Question 16
17 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW An event has occurred resulting in a reactor scram. The EOPs have been entered, and the following conditions currently exist:
- Reactor power is 15% and steady.
- SRVs are being used to maintain reactor pressure below 1093 psig.
- RPV water level is being maintained 0-50 inches.
Which ONE of the following conditions will allow the operating crew to restore RPV level to the normal band?
A. Reactor period is negative B. All IRMs indicate downscale C. SLC Tank level is <45 inches D. Two rods at position 04, all remaining rods at 02 Answer: C 2017 ILO Written Page: 34 of 207 Question 17
Answer Explanation:
<45 inches in the SLC Tank is the definition of Hot Shutdown Boron Weight, and override FSL-OR2 allows RPV level to be restored to the normal band.
Distractor Explanation:
If the reactor is shutdown under ALL conditions, Sheet 1 can be entered which restores RPV level to the normal band. All distracters are plausible because they could be used to decide the reactor is shutdown.
However, none of the indications are valid determinations that the reactor is shutdown under all conditions.
A. Negative reactor period is not indicative that the reactor is shutdown under all conditions, only that power level is lowering.
B. IRM indication is not a valid means to determine the reactor is shutdown under all conditions.
D. All rods <02, all control rods except one <02, OR Cold Shutdown Boron Weight injected, are required to exit the ATWS side and restore RPV level. Two control rods >02 do not meet the definition.
Reference Information:
EPG (B-14-43 to 45) 29.100.01 SH 1A (FLS-OR2)
Plant Procedures 29.100.01 SH 1A BWROG EPG App B NUREG 1123 KA Catalog Rev. 2 295037 EK3. Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :
295037 EK3.04 Hot shutdown boron weight: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Low New RO 2017 ILO Written Page: 35 of 207 Question 17
18 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: NEW An offsite release is in progress with CCHVAC running in Recircualtion mode. If both the North and South Emergency Makeup intakes recieve a Hi-Hi radiation signal while the intake selector switch is in AUTO, how will Emergency Makeup logic respond?
A. Both inlets will remain open.
B. Both inlets will close for a five minute sampling period.
C. Both inlets will close and remain closed.
D. Both inlets will open for a 5 minute sampling period.
Answer: A or D 2017 ILO Written Page: 36 of 207 Question 18
Answer Explanation:
If both of the radiation monitors are both above or below the Hi-Hi setpoint the logic will not select either intake and both the intakes will remain open. The operator must select the Emergency intake to open using radiation level indication in the Relay Room. This will occur With the 3 position switch for Emergency air intake selected to AUTO.
Distractor Explanation:
Distractor are plausible and incorrect because the Air intakes take a 5 minute sample but both intakes remain open during the sampling.
Reference Information:
ST-OP-315-0073 (pg 20) - Training material I-2611-51 (T41M72) relay that forces both open - Technical Reference NUREG 1123 KA Catalog Rev. 2 295038 EA1. Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE :
295038 EA1.07 Control room ventilation: Plant-Specific NRC Exam Usage ILO 2017 Exam ILO 2017 NRC Question Use:
Low New RO 2017 ILO Written Page: 37 of 207 Question 18
19 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: BANK The plant was operating at 100% power when the following alarms were received:
- 16D27, FIRE ALARM
- 7D10, ELECTRIC FIRE PUMP AUTO START In addition to sounding the Fire Alarm and announcing the event, which ONE of the following lists operator actions that are IMMEDIATELY required?
A. Notify Fire Brigade Leader and activate Fire Brigade.
B. Dispatch an operator to investigate and inform Security.
C. Notify Fire Brigade Leader and inform Security of magnitude, location, and type of fire.
D. Activate Fire Brigade, establish communication between Control Room and Fire Brigade, and call Frenchtown Fire Dept.
Answer: A 2017 ILO Written Page: 38 of 207 Question 19
Answer Explanation:
Immediate operator actions for a confirmed fire per 20.000.22 are: Sound Fire Alarm, Announce event over Hi-Com, Notify Fire Brigade Leader, and Activate Fire Brigade. Receipt of 7D10 is a secondary indication of a confirmed fire.
Distractor Explanation:
B. Is plausible and incorrect because dispatching an operator to investigate is a subsequent action for a fire alarm received with no confirmation.
C. Is plausible and incorrect because informing Security of magnitude, location and type of fire is a subsequent action.
D. Is plausible and incorrect because establishing communication between Control Room and Fire Brigade is a subsequent action.
Reference Information:
20.000.22 (pg 3)
Plant Procedures 07D10 16D27 20.000.22 NUREG 1123 KA Catalog Rev. 2 600000 Plant Fire On Site G2.1.20 Ability to interpret and execute procedure steps 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank High RO 2017 ILO Written Page: 39 of 207 Question 19
20 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant is operating at 80% power during a rod pattern adjustment. The Systems Operations Center notifies the main control room that there is potential for degraded grid conditions due to erratic reactive loading in Monroe County.
4D132, Generator Frequency High/Low, alarms, and the CRLNO reports that frequency is currently indicating 58 Hz.
Which of the following operator actions is required, and why?
A. Reduce generator load to raise frequency to stabilize the grid.
B. Start CTG 11-1 to minimize the impact of a potential loss of the 120Kv Mat.
C. Scram the reactor to protect the plant from conditions that might damage the Main Generator.
D. Start EDG 11 and EDG 12 to protect safety related equipment from possible low voltage high current conditions.
Answer: C 2017 ILO Written Page: 40 of 207 Question 20
Answer Explanation:
With generator frequency less than 58.2% and power greater than 30% (conditions which could damage the generator), 20.300.GRID, Override 1, directs placing the mode switch in Shutdown to protect the plant.
Distractor Explanation:
A. Is plausible and incorrect because Fermi operations may not take independent action to stabilize the grid. SOC/ITC must coordinate those actions.
B. Is plausible and incorrect because is not directed by the 20.300.GRID.
D. Is plausible and incorrect because this is directed only if SST 64 Bus alarms on low voltage.
Reference Information:
20.300.GRID, page 2 Plant Procedures 20.300.GRID 20.300.GRID BASES NUREG 1123 KA Catalog Rev. 2 700000 AK3. Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:
700000 AK3.01 Reactor and turbine trip criteria 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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High New RO 2017 ILO Written Page: 41 of 207 Question 20
21 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: MODIFIED The plant is operating at 100 % power, steady state, when Condenser Circulating Water System malfunctions cause condenser vacuum to commence degrading. The crew enters 20.125.01 Loss Of Condenser Vacuum and starts performing actions to mitigate the event. The following conditions exist currently:
- SJAE all in service
- Condenser Vacuum 1.45 psia (degrading)
- Reactor Power 90% and lowering.
As condenser vacuum continues to degrade from these conditions, which of the following condenser pressure changes would result in a main turbine automatic trip with the Main Turbine Bypass Valves NOT tripped?
A. + 1 psi B. + 2 psi C. + 4 psi D. + 6 psi Answer: C 2017 ILO Written Page: 42 of 207 Question 21
Answer Explanation:
Turbine trip setpoint (3.68 psia) minus current condition (1.45 psia) = 2.23 psi. (6.8 psia) Main Turbine Bypass Valve Closure minus current condition (1.45 psia) = 5.35 psi. An increase of 4 psi would cause a turbine trip and NOT trip the Main Turbine Bypass Valves Distractor Explanation:
Answers are plausible and incorrect based on setpoints and calculations provided in the Answer Explanation.
Reference Information:
20.125.01 NOTE 1 pg 2 Plant Procedures 04D108 20.125.01 NUREG 1123 KA Catalog Rev. 2 295002 AK2. Knowledge of the interrelations between LOSS OF MAIN CONDENSER VACUUM and the following:
295002 AK2.02 Main turbine 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Low Modified RO 2017 ILO Written Page: 43 of 207 Question 21
22 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK A seismic event has occurred during a plant startup. A Group I Isolation occurred while at 8%
power. RPS has initiated a scram on MSIV closure. No other automatic RPS trips have actuated and no operator actions have been taken.
At what pressure will the first SRV open?
A. 1017 psig B. 1135 psig C. 1047 psig D. 1093 psig Answer: B 2017 ILO Written Page: 44 of 207 Question 22
Answer Explanation:
Per 23.201:
Low Low Set logic is activated, when both signals have sealed in; any SRV open and Reactor pressure greater than high Reactor Vessel Steam Dome Pressure Scram setpoint of 1093 psig. The logic opens and closes B2104-F013A, Div 1 MS Line D SRV, and B2104-F013G, Div 2 MS Line B SRV, so Reactor pressure is controlled automatically. SRV A controls Reactor pressure between 905 and 1017 psig, and SRV G controls Reactor pressure between 935 and 1047 psig.
Since the first SRV opens at 1135 psig (per enclosure A, 23.201), No SRV will open until then. After that, LLS will be armed and cycle to control pressure.
Distractor Explanation:
The distractors are plausible because they are pressures that relate to SRV operations. (LLS and Hi Pressure)
Reference Information:
23.201 Encl A Plant Procedures 23.201 NUREG 1123 KA Catalog Rev. 2 295025 EK3. Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE :
295025 EK3.01 Safety/relief valve opening\
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (3) Mechanical components and design features of reactor primary system.
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Bank Low RO 2017 ILO Written Page: 45 of 207 Question 22
23 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW The plant is at 100% when RWL increases to the L8 Trip setpoint. Which ONE of the following describes the plant response and the reason for the response.
The L8 instruments directly cause __(1)__ and this will __(2)__
A. (1) Reactor Scram (2) mitigate the reduction in MCPR caused by the L8.
B. (1) Reactor Scram (2) prevent the erosion of the Main Turbine Blades.
C. (1) Turbine Trip.
(2) mitigate the reduction in MCPR caused by the L8.
D. (1) Turbine Trip.
(2) prevent the erosion of the Main Turbine Blades.
Answer: D 2017 ILO Written Page: 46 of 207 Question 23
Answer Explanation:
Per 23.601 Reactor Vessel High Water Level - Level 8, (A or B) and (C or D) = Main Turbine/N and S FW Turbine Trip.
Per TSB 3.3.2.2 A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine. In this case the TSB are the reference document for the information provided in student text and training material for reason for the Level 8 Main Turbine Trip.
Distractor Explanation:
Per TSB 3.3.2.2 The Level 8 trip indirectly initiates a reactor scram (above 30% RTP) from the main turbine trip and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR (caused by the Turbine Trip vs the L8). Based on the stem of the question this is incorrect in the context of the question.
Reference Information:
23.601 Pg 22 TSB 3.3.2.2 (Background & Safety Analysis)
ILO 2017 Exam Note:
Validation data SRO 80% pass Validation data RO 70% pass LOD of 2 discussion. Per the DBD for RPS the indirect trip that comes from the limit switches and solenoid on the turbine valves does benefit the margin to MCPR, however this is not the reason for the trip. The trip is for the Turbine and is sent to the Turbine valves. The resulting SCAM is a chained event that happens because of what the turbine valves do. This is a Trip and bases question, however there is a complexity to this question because of the relationship with RPS. Validation data and this complexity makes this a LOD 2 question Plant Procedures 23.601 NUREG 1123 KA Catalog Rev. 2 295008 AK1. Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR WATER LEVEL :
295008 AK1.01 Moisture carryover Technical Specifications 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.
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Low New RO 2017 ILO Written Page: 47 of 207 Question 23
24 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: BANK The plant is operating at 70% power when multiple alarms are received, and the following conditions are noted:
- RPV level 192 inches (decreasing)
- N RFPT speed increasing
- S RFPT speed decreasing
- Steam Flow/Feed Flow Recorder steam flow > feed flow Both SBFW pumps are started and injecting at rated flow. RPV level is now 178 inches and continues to decrease.
Given these conditions, which of the following actions will be taken in accordance with plant procedures?
A. Place Mode Switch in Shutdown.
B. Manually initiate and inject with HPCI.
C. Manually initiate and inject with RCIC.
D. Place the Recirculation System A & B Flow Limiter 2/3 Defeat Switch in DEFEAT.
Answer: A 2017 ILO Written Page: 48 of 207 Question 24
Answer Explanation:
This combination of RFPs and SBFW pumps is not recovering level. Per immediate actions 20.107.01 A1, Place mode switch in SHUTDOWN.
Distractor Explanation:
B. Is incorrect plausible because HPCI is not within the bounds of the analysis for increased subcooling. It is not the approved method of water addition for the situation given.
C. Is incorrect plausible because there is insufficient time to turn the transient before a L3 SCRAM. It takes the same amount of time to start and inject with RCIC as it does with SBFW. Loss of 14" is an indicator that L3 would occur before RCIC could beplaced in service and turn level.
D. Is incorrect and plausible because the 2/3 defeat should be placed in NORMAL per 20.107.01 Reference Information:
20.107.01 Immediate actions.
Plant Procedures 20.107.01 NUREG 1123 KA Catalog Rev. 2 295009 AA1. Ability to operate and/or monitor the following as they apply to LOW REACTOR WATER LEVEL :
295009 AA1.01 Reactor feedwater 295009 AA2. Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL :
295009 AA2.01 Reactor water level 295009 AA2.02 Steam flow/feed flow mismatch 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank High RO Selected for Early Review (NRC) 2017 ILO Written Page: 49 of 207 Question 24
25 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is operating at full power and the following conditions exist:
- 14 DW Cooling Fans are operating.
- T47-R803A, Drywell Cooling System Area Temperatures Div 1, point 16 indicates > 175°F (alarm setpoint is 155°F).
- Lake Temperature is 71°F.
The AVERAGE Drywell Temperature has risen from 132°F to 136°F during the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and the Primary Containment Pressure Recorder indicates Drywell Pressure has risen from 14.4 inWC to 16.6 inWC.
Which ONE of the following actions is appropriate?
A. Operate ALL available Drywell Cooling per 29.100.01.
B. Place RBCCW SCS in service in accordance with 23.127.01.
C. Manually INITIATE EECW and EESW Systems per 20.127.01.
D. Increase cooling water flow using P42-F400, RBCCW Temp Control Vlv, in AUTO or MANUAL.
Answer: B 2017 ILO Written Page: 50 of 207 Question 25
Answer Explanation:
With High DW temperature and RBCCW TCV fully open, it is required to place RBCCW Supplemental Cooling into service.
Distractor Explanation:
A. The EOP entry condition (145F) has not been exceeded C. Is appropriate if there is a loss of RBCCW.
D. The RBCCW temperature control valve (TCV) is inline and already at 100%. However if the examinee incorrectly assumes the TCV is a bypass valve they would believe that closing the TCV would increase cooling flow.
Reference Information:
ARP 8D41 (pg 1)
Plant Procedures 08D41 23.127.01 NUREG 1123 KA Catalog Rev. 2 295012 AK1. Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE :
295012 AK1.01 Pressure/temperature relationship 295012 AK3. Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE :
295012 AK3.01 Increased drywell cooling 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Bank High RO 2017 ILO Written Page: 51 of 207 Question 25
26 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK Reactor pressure is 800 psig during a reactor startup. Following a CRD Pump trip, the 3D10 CRD ACCUMULATOR TROUBLE alarm is received on a withdrawn control rod. What action is required, and is the reason for this action?
A. Place Mode Switch in Shutdown IMMEDIATELY. This is a conservative measure; because by design, control rod scram can only be accomplished without the accumulator pressure when the reactor vessel pressure is at or above 600 psig.
B. Place Mode Switch in Shutdown IMMEDIATELY. This is a conservative measure; because by design, control rod scram can NOT be accomplished without the accumulator pressure when the reactor vessel pressure is at or above 600 psig C. Place Mode Switch in Shutdown in 20 Minutes. The allowed completion time of 20 minutes is reasonable; to place a CRD pump into service to restore the charging header pressure.
D. Place Mode Switch in Shutdown in 20 Minutes. All of the accumulators would not have sufficient pressure, resulting in a potentially severe degradation of the scram performance.
Answer: A 2017 ILO Written Page: 52 of 207 Question 26
Answer Explanation:
Per AOP and BASES, anytime below 900 psig a CRD Accumulator trouble light is on for a withdrawn rod, the mode switch is placed in shutdown. This is a conservative measure because by design, control rod scram can be accomplished without the accumulator pressure when the reactor vessel pressure is at or above 600 psig Distractors Explanation:
B. is incorrect and plausible because of the basis.
C. is incorrect and plausible because but would be true if pressure was greater than 900 psig.
D. is incorrect and plausible because but would be true if pressure was greater than 900 psig.
Reference Information:
20.106.01 Basis Override 1 (pg2)
Plant Procedures 20.106.01 NUREG 1123 KA Catalog Rev. 2 295022 AK2. Knowledge of the interrelations between LOSS OF CRD PUMPS and the following:
295022 AK2.03 Accumulator pressures 295022 AK3. Knowledge of the reasons for the following responses as they apply to LOSS OF CRD PUMPS:
295022 AK3.01 Reactor SCRAM 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Bank High RO 2017 ILO Written Page: 53 of 207 Question 26
27 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK The following actions automatically occurred as the result of High Radiation:
- Reactor Building HVAC tripped.
- SGTS started.
- Control Center HVAC aligned to recirculation mode.
Based on the information provided above, which ONE of the following radiation monitor indications is the LOWEST value that can cause these automatic actions?
A. Fuel Pool Vent Exhaust indicating 7.3 mRem/hr.
B. Reactor Building Vent Exhaust indicating 14,500 cpm.
C. Fuel Pool Vent Exhaust indicating 2.3 mRem/hr.
D. Reactor Building Vent Exhaust indicating 11,500 cpm.
Answer: A 2017 ILO Written Page: 54 of 207 Question 27
Answer Explanation:
Per ARP 3D35 Fuel Pool Vent Exhaust above 3mr/hr will cause these automatic isolations and actuations.
Distractor Explanation:
Reactor Building Vent Exhaust is incorrect and plausible because the trip setpoint is 16,000 cpm Fuel Pool Vent Exhaust indicating 2.3 is incorrect and plausible because Answer Explanation.
Reference Information:
ARP 3D35 (pg 2)
Plant Procedures 03D035 NUREG 1123 KA Catalog Rev. 2 295034 EK2. Knowledge of the interrelations between SECONDARY CONTAINMENT VENTILATION HIGH RADIATION and the following:
295034 EK2.04 Secondary containment ventilation G2.1.27 Knowledge of system purpose and or function.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Bank Low RO 2017 ILO Written Page: 55 of 207 Question 27
28 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant is operating at 100% power All RPV water level indication is lost.
The crew has entered 29.100.01 SH 3A RPV flooding - ATWS 5 SRVs are open RHR is the only system available for injection.
Which one of the following is the highest pressure when the CRS would order starting a RHR pump to provide adequate core cooling (ACC) and why?
A. 310 psig becaused this is the highest discharge pressure the LPCI pump can develop.
B. 285 psig becaused this is the highest discharge pressure the LPCI pump can develop.
C. 225 psig because below this pressure LPCI injection is required.
D. 100 psig because below this pressure SRVs will be closed.
Answer: C 2017 ILO Written Page: 56 of 207 Question 28
Answer Explanation:
Per MSCP with 5 SRV open pressure must be less than 230. EOP Step FSRF-4 Distractor Explanation:
A. Is incorrect and plausible because this is the hight discharge pressure the LPCI pump can develop (29.ESP.01) however per answer explaination this pressure is too high.
B. Is incorrect and plausible because this is the hight discharge pressure the CS pump can develop (29.ESP.01) however per answer explaination this pressure is too high.
D. Is incorrect and plausible because it is not the HIGHEST pressure where injection is permissable per the EOPs Reference Information:
EOP 29.100.01 SH 3A Plant Procedures 29.100.01 SH 3A NUREG 1123 KA Catalog Rev. 2 203000 K3. Knowledge of the effect that a loss or malfunction of the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) will have on following:
203000 K3.04 Adequate core cooling 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High New RO 2017 ILO Written Page: 57 of 207 Question 28
29 Points: 1.00 Difficulty: 4.00 Level of Knowledge: High Source: MODIFIED RHR Loop B is in Shutdown Cooling, and RPV Water level is 194 inches and lowering. A maintenance mishap has caused a line to be sheared. The line is the reference line for the following instruments:
- B21-N080C DIV 1 REACTOR LEVEL NARROW RANGE TRANSMITTER.
- B21-N080D DIV 1 REACTOR LEVEL NARROW RANGE TRANSMITTER.
Shortly after this incident ACTUAL RPV level begins lowering at 2 inches a minute.
Assuming no operator action what is the position of the following valves 15 minutes after ACTUAL RPV level started to lower?
E1150-F008 E1150-F009 RHR SDC OTBD RHR SDC INBD SUCTION ISO VLV SUCTION ISO VLV A. OPEN OPEN B. CLOSE OPEN C. OPEN CLOSE D. CLOSE CLOSE Answer: Question Deleted 2017 ILO Written Page: 58 of 207 Question 29
Answer Explanation:
Per M-2090 B21-N080C/D are supplied from a single tap. Failure to the reference leg will make a level instrument fail HIGH. This means that for NSSSS (C & D) logic will never be met. 23.601 Trip sheet shows that A&C will be met at below 173.4 inches. 2 inches per min for 15 minutes is 30 inches. 194-30
= 164 with is less than 173.4. A&C NSSSS closes the E1150-F009 Distractor Explanation Distractor are incorrect and plausible based on Answer Explanation.
Reference Information:
M-2090 23.601 (pg 11)
Plant Procedures 23.601 Maintenance Plant Drawings M-2090 NUREG 1123 KA Catalog Rev. 2 205000 A1. Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including:
205000 A1.05 Reactor water level 205000 K4. Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following:
205000 K4.03 Low reactor water level: Plant-Specific 205000 K6. Knowledge of the effect that a loss or malfunction of the following will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) :
205000 K6.04 Reactor water level 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High Modified RO 2017 ILO Written Page: 59 of 207 Question 29
30 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED The plant has been shutdown for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following 300 days of power operation with the following conditions:
- RHR Pump B is operating in Shutdown Cooling Mode.
- RPV Level is 230 inches.
Which of the following actions should be taken for an over-current trip of RHRSW Pump B?
A. Throttle closed E1150-F068B, Div 2 RHR HX Serv WTR Outlet FCV, to establish proper RHRSW flow AND monitor and record RPV temperature and pressure IAW 22.000.05, Pressure/Temperature Monitoring During Heatup And Cooldown" B. Throttle open E1150-F068B, Div 2 RHR HX Serv WTR Outlet FCV, to maximize service water flow AND monitor and establish decay heat removal per 23.800.05, Alternate Remote Coolant Circulation and Decay Heat Removal.
C. Fully close E1150-F048B, RHR Heat Exchanger Bypass Valve, to increase heat exchanger load AND monitor and establish decay heat removal per 23.800.05, Alternate Remote Coolant Circulation and Decay Heat Removal.
D. Fully open E1150-F048B, RHR Heat Exchanger Bypass Valve, to reduce heat exchanger load AND monitor and record RPV temperature and pressure IAW 22.000.05, Pressure/Temperature Monitoring During Heatup And Cooldown" Answer: A 2017 ILO Written Page: 60 of 207 Question 30
Answer Explanation:
RHRSW valve is throttled closed to meet pump DP requirements. Monitor and record RPV temperature and pressure IAW 22.000.05, Pressure/Temperature Monitoring During Heatup And Cooldown" is required by SOP for SDC Distractor Explanation:
B. Is incorrect and plausible because Throttle open E1150-F068B will increase flow and cooling lost by pump trip, however it is NOT allowed by procedure because flow must be REDUCED per NOTE in SOP to a range of 5400 gpm to 6300 gpm to prevent excessive vibration of E1150-F068B, Div 1 RHR Hx Serv Wtr Outlet FCV. 23.800.05, Alternate Remote Coolant Circulation and Decay Heat Removal COULD be used and is directed from the loss of SDC AOP.
C. Is incorrect and plausible because FULLY closing E1150-F048B would greatly increase cooling to the fuel and cooling has been already reduced with the loss of the RHRSW pump, for this event it would be good to throttled in the closed direction to maintain cooldown or the the current pressure/temperature band HOWEVER the SOP specifically precludes FULLY closing the E1150-F048B when ONLY ONE RHR pump is running in the loop. Monitor and record RPV temperature and pressure IAW 22.000.05, Pressure/Temperature Monitoring During Heatup And Cooldown" is required by SOP for SDC D. Is incorrect and plausible because FULLY opening E1150-F048B would greatly reduce cooling when cooling has aready been reduced. 23.800.05, Alternate Remote Coolant Circulation and Decay Heat Removal COULD be used and is directed from the loss of SDC AOP.
Reference Information:
23.205 Section 6.2 (SDC Div 2 RHR) 23.208 Section 5.4 (RHRSW operation)
Plant Procedures 23.205 23.208 NUREG 1123 KA Catalog Rev. 2 205000 A4. Ability to manually operate and/or monitor in the control room:
205000 A4.04 Heat exchanger cooling water valves 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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High Modified RO Selected for Early Review (NRC) 2017 ILO Written Page: 61 of 207 Question 30
31 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: BANK Which of the following responses best completes the statement below regarding a limitation associated with the operation of the HPCI system?
When the HPCI turbine is running, its speed must be maintained above __(1)__ rpm to ensure
__(2)__.
A. (1) 2000 (2) adequate discharge pressure to allow injection into the RPV B. (1) 2000 (2) stability of operation C. (1) 4000 (2) stability of operation D. (1) 4000 (2) adequate discharge pressure to allow injection into the RPV Answer: B 2017 ILO Written Page: 62 of 207 Question 31
Answer Explanation:
As stated in the precautions and limitations in 23.202.
Distractors Explanation:
A. Is incorrect and plausible because HPCI could be at lower RPM depending on what RPV pressure is and for testing.
C. Is incorrect and plausible rpm is wrong .
D. Is incorrect and plausible because rpm is wrong with incorrect limitation Reference Information:
23.202 P&L 3.3 Plant Procedures 23.202 NUREG 1123 KA Catalog Rev. 2 206000 K5. Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM :
206000 K5.01 Turbine operation: BWR-2,3,4 206000 K5.05 Turbine speed control: BWR-2,3,4 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Bank Low RO 2017 ILO Written Page: 63 of 207 Question 31
32 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW With the plant operating at 100% power, which of the following sets of conditions requires immediate entry into LCO 3.0.3?
A. HPCI System inoperable.
B. Two or more ADS valves inoperable.
C. HPCI and one or more ADS valves inoperable.
D. One LPCI pump in both LPCI subsystems inoperable.
Answer: C 2017 ILO Written Page: 64 of 207 Question 32
Answer Explanation:
Per 3.5.1 Condition K, HPCI and one or more ADS valves inoperable enter LCO 3.0.3 immediately.
Distractor Explanation:
A. Is incorrect and plausible because it is Condition E, and does not required LCO 3.0.3 B. Is incorrect and plausible because it is Condition J, and does not required LCO 3.0.3 D. Is incorrect and plausible because it is Condition B, and does not required LCO 3.0.3 Reference Information:
T.S. 3.5.1 NUREG 1123 KA Catalog Rev. 2 206000 HPCI System.
G2.2.39 Knowledge of less than one hour technical specification action statements for systems.
Technical Specifications 3.5.1 ECCS Operating 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Low New RO 2017 ILO Written Page: 65 of 207 Question 32
33 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: MODIFIED While operating at 100% power, a seismic event causes a loss of all offsite power AND a recirculation loop rupture.
If EDG # 11 fails to start, what would be the impact of this electrical loss on low pressure injection?
A. Core Spray Pump A is de-energized.
B. Core Spray Pump B is de-energized.
C. Core Spray Pump C is de-energized.
D. Core Spray Pump D is de-energized.
Answer: A 2017 ILO Written Page: 66 of 207 Question 33
Answer Explanation:
Only Core Spray Pump A is powered from 4160v Bus 64B.
Distractor Explanation:
B. Is plausible and incorrect, CS pump B is powered from 4160v Bus 65E C. Is plausible and incorrect, RHR pump C is powered from 4160v Bus 64C D. Is plausible and incorrect, RHR pump D is power from 4160v Bus 65F Reference Information:
I-2215-02 (G-7) Power Monitor for Pump A Plant Procedures 23.203 NUREG 1123 KA Catalog Rev. 2 209001 K2. Knowledge of electrical power supplies to the following:
209001 K2.01 Pump power 209001 K6. Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM :
209001 K6.02 Emergency generators 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Low Modified RO 2017 ILO Written Page: 67 of 207 Question 33
34 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW The plant is shutdown in MODE 4 and DIV 1 Core Spray is in operation for surveillance testing.
RHR Pump B is in service for shutdown cooling.
Flow through DIV 1 Core Spray is being controlled by E2150-F015A, Div 1 CS Test Line Iso Valve at 3500 gpm.
Reactor water level suddenly lowers.
If Reactor water level continues to lower to __(1)__ inches the non-running core spray pumps will start and the E2150-F015A, Div 1 CS Test Line Iso Valve will __(2)__.
A. (1) 31.8 (2) remain OPEN.
B. (1) 31.8 (2) CLOSE C. (1) 110.8 (2) remain OPEN.
D. (1) 110.8 (2) CLOSE Answer: B 2017 ILO Written Page: 68 of 207 Question 34
Answer Explanation:
Per 23.203, Section 5.5, Auto Initiation Div 1, "If Div 1 Core Spray is in the Test Mode when an automatic initiation signal is received, E2150-F015A, Div 1 CS Test Line Iso Vlv, will close as the system aligns for injection to the Reactor Vessel." Also shown on I-2210-01 that L1 or High Drywell will start Core Spray Pumps and close the E2150-F015A.
Distracter Explanation:
110.8 inches is plausible and incorrect because it is the L2 setpoint, and the E2150-F015A is plausible and incorrect if the examinee does not understand the setpoints associated with actuations and isolations CS logic.
Reference Information:
I-2210-01 Logice Diagram Core Spray 23.203 Section 5.5 (pg 22) NOTE I-2215-02 Core Spray Logic 23.601 Pg 16 Reactor Vessel Low Water Level - Level 1 Plant Procedures 23.203 NUREG 1123 KA Catalog Rev. 2 209001 A1. Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including:
209001 A1.03 Reactor water level 209001 A1.08 System lineup 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Low New RO 2017 ILO Written Page: 69 of 207 Question 34
35 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK During an ATWS, SLC Pump B has been started. The following indications are observed:
- RPV Pressure is 1015 psig.
- SLC Tank Level is 72 inches decreasing.
- SLC Pump Discharge Pressure is 1050 psig.
- 3D11, SLC IGNITION CONTINUITY LOSS, is activated.
- BOTH H11-P603 Continuity Circuit indicating lights are OFF.
- G3352-F001, RWCU Supply Inbd Iso Vlv, remains OPEN.
The above conditions indicate which of the following?
A. Normal operation.
B. SLC Explosive Valves failed to fire.
C. RWCU failed to isolate on SLC injection.
D. SLC Pump B Discharge Relief Valve failed open.
Answer: A 2017 ILO Written Page: 70 of 207 Question 35
Answer Explanation:
Proper indication of SLC injection include RWCU isolation (G3352-F004 and F220 only), SLC tank level decreasing, 3D11 activated, Continuity Circuit indicating lights off, and pressure slightly higher than RPV pressure.
Distractor Explanation:
B. Is incorrect and plausible because if they failed to fire, 3D11 would not alarm, Continuity Circuit indicating lights would be on, and pressure would be higher.
C. Is incorrect and plausible because RWCU did isolate properly, but there is a common misconception G3352-F001 shuts on SLC injection.
D. Is incorrect and plausible because even though pressure is higher, it is not fluctuating as indication of relief valve stuck open.
Reference Information:
23.139, page 11 Plant Procedures 23.139 NUREG 1123 KA Catalog Rev. 2 211000 A1. Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including:
211000 A1.10 Lights and alarms 211000 A4. Ability to manually operate and/or monitor in the control room:
211000 A4.07 Lights and alarms G2.1.31 Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Bank Low RO 2017 ILO Written Page: 71 of 207 Question 35
36 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: BANK An operator standing by a Hydraulic Control Unit (HCU) bank hears a click of multiple solenoid valves changing position, and no sound of air discharging. This is followed by an announcement from the Control Room of the receipt of a 1/2 scram.
Which ONE of the following is the initiating event which leads to this condition?
A. A 1/2 scram of RPS Channel A.
B. A 1/2 scram of RPS Channel B.
C. A 1/2 scram caused one half of the Back-Up Scram Valves to energize.
D. A 1/2 scram caused one half of the Back-Up Scram Valves to de-energize.
Answer: A 2017 ILO Written Page: 72 of 207 Question 36
Answer Explanation:
When RPS A cause a half scram, solenoids will reposition, no air would vent.
Distractor Explanation:
B. is incorrect for if RPS B initiated the half scam, air would be vented.
C/D. is incorrect because backup scam valves energize when full scram signal is present.
Reference Information:
M-5703-1 Maintenance Plant Drawings m-5703-1 NUREG 1123 KA Catalog Rev. 2 212000 K1. Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following:
212000 K1.06 Control rod drive hydraulic system 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (6) Design, components, and function of reactivity control mechanisms and instrumentation.
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Bank Low RO 2017 ILO Written Page: 73 of 207 Question 36
37 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK The reactor mode switch is in REFUEL and the shorting links are installed. IRM G Hi Hi Upscale trip is observed. What is the response of the RPS and Reactor Manual Control Systems?
A. HALF Scram: No Rod Block.
B. FULL Scram: No Rod Block.
C. HALF Scram: Rod Block.
D. FULL Scram: Rod Block.
Answer: C 2017 ILO Written Page: 74 of 207 Question 37
Answer Explanation:
With shorting links installed IRM A/E/C/G High flux (120 on 125 scale) causes A1/A2 of RPS to trip, causing half scram. Additionally with the mode switch in refuel IRM upscale causes a Rod block.
Distracter Information:
Half Scram and No Rod block are incorrect and plausible based on referenced prints Reference Information:
I-2115-6 Shows how mode switch effects rod blocks from SRM/IRM ARP 3D73 (pg 2) REACTOR SCRAM
SUMMARY
ARP 3D11 (pg 2) ROD BLOCK
SUMMARY
Plant Procedures 03D011 03D073 NUREG 1123 KA Catalog Rev. 2 215003 K1. Knowledge of the physical connections and/or cause effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following:
215003 K1.01 RPS 215003 K1.02 Reactor manual control 215003 K4. Knowledge of INTERMEDIATE RANGE MONITOR (IRM) SYSTEM design feature(s) and/or interlocks which provide for the following:
215003 K4.01 Rod withdrawal blocks 215003 K4.02 Reactor SCRAM signals 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (6) Design, components, and function of reactivity control mechanisms and instrumentation.
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Bank Low RO 2017 ILO Written Page: 75 of 207 Question 37
38 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: MODIFIED During a reactor startup, reactor power is on Range 3 of the Intermediate Range Monitors, and Source Range Monitor (SRM) detectors are being withdrawn from the core.
DIV 1 48/24V DC Distribution loses power.
Assuming no operator action is taken, a Control Rod Block is caused by power loss to _____ .
D. SRMs B, D and IRMs B,D,F,H.
Answer: C 2017 ILO Written Page: 76 of 207 Question 38
Answer Explanation:
Power to IRM Channels A, C, E, G and SRM Channels A and C is from 48/24V DC Distribution Cabinet 2IA1-3 Distracter Explanation:
Distracters are plausible as SRM drives are powered from 120/208V Distribution Cabinet 72E-2B-1 vice 48/24V and knowing the ROD BLOCK setpoint.
References Information:
ARP 3D113, ROD BLOCK 300 V DC decreasing (pg. 2),
SOP 23.602, ENERGIZING SRM (Pg 9).
Plant Procedures 03D113 23.602 NUREG 1123 KA Catalog Rev. 2 215004 K2. Knowledge of electrical power supplies to the following:
215004 K2.01 SRM channels/detectors 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Low Modified RO 2017 ILO Written Page: 77 of 207 Question 38
39 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED A plant startup is in progress. Reactor power is 30% and being increased by control rod withdrawal. APRM 1 malfunctions and is indicating 0% power.
Which one of the following describes the effect on Rod Block Monitors as a result of the APRM failure?
A. RBM A and B are functional. Outward Control rod motion can continue.
B. RBM A and B are NOT functional. Outward Control rod motion is blocked.
C. RBM A is NOT functional. Bypassing APRM 1 will allow outward control rod motion.
D. RBM B is NOT functional. Bypassing APRM 1 will allow outward control rod motion.
Answer: C 2017 ILO Written Page: 78 of 207 Question 39
Answer Explanation:
The examinee should correctly recognize that APRM #1 is the primary APRM for RBM A and determine that, under the current condition of 30% power, the RBM is bypassed due to APRM 2 power input of 0%.
The RBM is bypassed until power exceeds the setpoint of 27% power from the primary APRM. The secondary APRM (APRM 4 in this case) will be automatically selected after the APRM 2 bypass joystick is operated.
Distracter Explanation:
A. Is incorrect and plausible because the examinee may incorrectly recall that technical specification allowable setpoint for RBM bypassing is 30% and determine that the RBM are not yet affected.
Additionally RBM A is associated with APRMs #1 and 3, and since it is may seem more logical to associate APRMs 1&2 with RBM A, rather than APRMs 2&4 being associated with RBM B. Also with the APRM downscale with the mode switch in run, outward rod motion could not occur due to a APRM downscale control rod withdrawl block.
B. Is incorrect but plausible because the examinee could incorrectly determine that the APRM downscale effects both RBMs. (Recirc flow transmitter failures could affect both RBMs flow comparator).
Outward control rod motion is blocked by the APRM downscale failure.
D. Is incorrect but plausible because the examinee may incorrectly recall that RBM B is associated with APRM #1.
Reference Information:
23.607 (pg 3) 1.1 System Description Plant Procedures 03D099 03D103 23.605 23.607 NUREG 1123 KA Catalog Rev. 2 215005 K3. Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following:
215005 K3.07 Rod block monitor: Plant-Specific Technical Specifications 3.3.2.1 Control Rod Block Instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High Modified RO 2017 ILO Written Page: 79 of 207 Question 39
40 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK Following a reactor scram and MSIV isolation, RPV water level dropped to 103 inches, then rose to 215 inches. Current RPV level is 182 inches (slowly lowering).
Given these conditions, which of the following actions are required to manually start RCIC?
A. Depress the Level 8 reset button, wait 15 seconds, and then open E5100-F095, RCIC Stm Inlet Bypass Valve.
B. Depress the Logic A and B Isolation reset button, and open E5100-F059, RCIC Turbine Inlet Trip Throttle Valve.
C. Open E5100-F095, RCIC Stm Inlet Bypass Valve, wait 15 seconds, and open E5100-F045, RCIC Steam Inlet Valve.
D. Close and re-open E5100-F059, RCIC Turbine Inlet Trip Throttle Valve, then open E5100-F045, RCIC Steam Inlet Valve.
Answer: C 2017 ILO Written Page: 80 of 207 Question 40
Answer Explanation:
Per 23.206, if RCIC is needed, the trip was due to L8 and RPV level is >L2, RCIC is to be started manually in accordance with section 5.4 "Manual Initition" The only actions that should be required (that match this question) is to Open E5100-F095, wait 15 seconds and then Open E5100-F045.
The distractors are plausible and depend on the examinees understanding of system operations and plant conditions.
Plant Procedures 23.206 NUREG 1123 KA Catalog Rev. 2 217000 K4. Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following:
217000 K4.06 Manual initiation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank High RO 2017 ILO Written Page: 81 of 207 Question 40
41 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED Plant is at 100% Power with annunciator 1D57, ADS/SRV/EECW TCV POWER SUPPLY FAILURE in alarm. Field investigation of 1D57 show that 2PA2-6 Circuit 1 is tripped.
A transient occurs resulting in conditions requiring the ADS system to function. How will ADS respond?
A. The ADS timer will start and all ADS valves will OPEN as designed.
B. The ADS timer will NOT start and the ADS valves will NOT open.
C. The ADS timer will start and all ADS valves will NOT open.
D. The ADS timer will NOT start, ADS valves will open at L1.
Answer: A 2017 ILO Written Page: 82 of 207 Question 41
Answer Explanation:
The examinee should also determine that ADS Logic String A has power; however, ADS Logic String B will have lost power from an alternate source (2PA2-6, Circuit 1).
Distracter Explanation:
Distractors are incorrect and plausible based on answer expainations.
Reference Information:
ARP 1D57 validates that some ADS components are de-energized due to 2PA2-6 Ckt 1 off as indicated in the stem The logic prints for ADS power supplies (I-2095-01) and B ADS logic power (I-2095-07) indicate how the logic remains automatically powered on a loss of the power from 2PA2-6 Ckt 1.
The logic prints for ADS valves P, J, R (I-2095-02) indicate how the valves remains automatically powered on a loss of the power from 2PA2-6 Ckt 1. Valves H and E are the same.
Plant Procedures 01D31 01D36 01D44 01D57 BWROG EPG App B NUREG 1123 KA Catalog Rev. 2 218000 K5. Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM:
218000 K5.01 ADS logic operation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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High Modified RO 2017 ILO Written Page: 83 of 207 Question 41
42 Points: 1.00 Difficulty: 4.00 Level of Knowledge: Low Source: NEW The plant is operating at power when it is discovered that an MTE instrument which failed post-use calibration was used to set all L1 instrumentation setpoints.
Which of the following includes ONLY Group isolations that would be impacted by this error?
Group 1 - MSIV & Drains Group 2 - Rx Sample Group 3 - RHR Group 4 - SDC/HD Spray Group 5 - CSS Group 6 - HPCI A. Groups 1,3,5 B. Groups 1,2,4 C. Groups 2,3,6 D. Groups 4,5,6 Answer: A 2017 ILO Written Page: 84 of 207 Question 42
Answer Explanation:
The automatic closure of various Class B valves prevents the release of significant radioactive material from the Primary Containment. These valves are not needed for a safe shutdown of the plant.
Distractor Explanation:
B. is incorrect and plausible; (2) Rx Sample + (4) SDC/HD Spray.
C. is incorrect and plausible; (6) HPCI D. is incorrect and plausible; (4) SDC/HD Spray + (6) HPCI Reference Information:
29.ESP.01 Pg 4 (under L1)
Plant Procedures 23.427 29.ESP.01 NUREG 1123 KA Catalog Rev. 2 223002 K6. Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF:
223002 K6.04 Nuclear boiler instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Low New RO 2017 ILO Written Page: 85 of 207 Question 42
43 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW Which of the following represent the response of Reactor Power and Main Turbine Generator Output, for a SRV opening for a few seconds and then closing?
A. #1 & #3 B. #1 & #4 C. #2 & #3 D. #2 & #4 Answer: B 2017 ILO Written Page: 86 of 207 Question 43
Answer Explanation:
Graphs show plant response. Both Reactor Power Power and Generator Output drop when a SRV is opened at 100% power. When the SRV closes, they return to normal.
Distractor Explanation:
Distactors are incorrect and plausible based on understanding the effect of SRV opening and closing during power operations.
Reference Information:
BWR Fundamentals NUREG 1123 KA Catalog Rev. 2 239002 A1. Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including:
239002 A1.07 Turbine load 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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High New RO 2017 ILO Written Page: 87 of 207 Question 43
44 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: BANK With reactor power at 100%, RFP pump speeds begin to oscillate. The CRLNO reviews the indications and with the agreement of the CRS takes the RFP B controller to EMERGENCY BYPASS mode. RFP B immediately stabilizes at its current value and A RFP speed returns to normal with no oscillations.
How will RPV level now respond and what actions are needed to restore level control?
A. Level lowers. Immediately lower recirc flow and insert CRAM array to allow capacity of feedwater to turn level.
B. Level lowers. Take manual control of RFP A and RFP B to raise Feed Flow and restore level.
C. Level will not change. Immediately lower recirc flow and insert CRAM array to lower feed demand to within the capacity of one RFP.
D. Level will not change. Continuously monitor for changes in RPV level until controller is repaired, control speed of RFP B if necessary in manual.
Answer: D 2017 ILO Written Page: 88 of 207 Question 44
Answer Explanation:
Level does not change. AOP 20.107.01 directs placing the affected RFP (B) to manual, and controlling speed as necessary to allow unaffected RFP (A) to control level.
Distractor Explanation:
A. Incorrect and plausible because RFP A will not lose signal so level will not lower.
B. Incorrect and plausible because RFP A will not lose signal so level will not lower.
C. Incorrect and plausible because incorrect action, current configuration is acceptable with operator monitoring level and controlling RFP B in manual with the controller in EMERGENCY BYPASS mode.
Reference Information:
20.107.01 & Basis Plant Procedures 20.107.01 23.107 NUREG 1123 KA Catalog Rev. 2 259002 A2. Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
259002 A2.06 Loss of controller signal output 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank Low RO 2017 ILO Written Page: 89 of 207 Question 44
45 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: MODIFIED Surveillance is being performed on Div 1 SGTS. The East and West trains of RBHVAC are in service.
Which of the following events would cause Division 1 SGTS flow as indicated by T46-R800A, Div 1 SGTS Exh Gas Flow Recorder to fall below 2000 scfm?
A. Div 1 SGTS 24kW Heater tripped.
B. SGTS Pre-Filter differential pressure is 1.3 water gauge.
C. Breaker for T4100-C004 RB East Exhaust Fan has tripped.
D. T46-F003A, Div 1 SGTS Exhaust Fan Inlet Vortex Damper, malfunctioning.
Answer: D 2017 ILO Written Page: 90 of 207 Question 45
Answer Explanation:
SGTS Exhaust Fan Inlet Vortex Damper modulates on start to maintain normal flow band. A malfunction could cause it to maintain a lower than normal flow.
Distractors Explanation:
A. Is incorrect and plausible because the heater will trip on alarm. If water collects in the charcoal it could restrict flow.
B. Is incorrect and plausible because it could lower flow, but vortex damper would maintain flow.
C. Is incorrect and plausible because RBHVAC has operating relationship with SGTS, however one train tripping will not cause flow to lower in running division of SGTS.
Reference Information:
8D35 Plant Procedures 08D35 23.404 NUREG 1123 KA Catalog Rev. 2 261000 A3. Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including:
261000 A3.03 Valve operation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Low Modified RO 2017 ILO Written Page: 91 of 207 Question 45
46 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED The plant is operating at 100% reactor power with EDG 13 currently synchronized to the GRID and being controlled from the Main Control Room.
A loss of the 120 Kv mat occurs.
Based on these conditions, which of the following statements correctly describes EDG operation and parameter response when controlled from the Main Control Room?
A. When EDG 12 Governor Control switch is taken to raise, EDG 12 Generator frequency will increase.
B. When EDG 12 EDG Voltage Control switch is taken to raise, EDG 12 Generator KVARS will increase.
C. When EDG 13 Governor Control switch is taken to raise, EDG 13 Generator frequency will increase.
D. When EDG 13 EDG Voltage Control switch is taken to raise, EDG 13 Generator Voltage will increase.
Answer: A 2017 ILO Written Page: 92 of 207 Question 46
Answer Explanation:
To answer this question the examinee must determine that EDG 13 would be in Droop mode and EDG 12 would be in isochronous. Understanding this is critical in controlling the EDG and being able to keep it's parameter in band as an operator. Based on this the Governor control switch and the Voltage control switch effect the EDG differently. For DROOP, Governor Control controls LOAD in KW and Voltage Control controls reactive load in VARS. For ISOCHRONOUS Governor Control controls frequency in Hz and Voltage Control controls output voltage.
Distracter Explanation:
B. is plausible and incorrect because it would be true if EDG 12 was in isochronous.
C. is plausible and incorrect because it would be true if EDG 13 was in droop.
D. is plausible and incorrect because it would be true if EDG 13 was in droop.
Reference Information:
ST-OP-315-0065-001 (pg 26)
Plant Procedures 23.307 NUREG 1123 KA Catalog Rev. 2 262001 A4. Ability to manually operate and/or monitor in the control room:
262001 A4.05 Voltage, current, power, and frequency on A.C. buses 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High Modified RO 2017 ILO Written Page: 93 of 207 Question 46
47 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: NEW A fire has damaged UPS A and because of the fire all loads for UPS A are denergized. UPS B continues to function normally.
Which of the following control room indications are impacted?
A. Rod Worth Minimizer (C11-J601)
B. Feedwater Line B Flow (C32-R604B)
C. Reactor Pressure Recorder (C32-R609)
D. RPIS indications on the Full Core Display (C1109-Y001)
Answer: D 2017 ILO Written Page: 94 of 207 Question 47
Answer Explanation:
Per ARP 3D22, RPIS indications on the Full Core Display are lost with UPS A.
Distractor Explanation:
Listed distractors are plausible and incorrect because they are power from UPS B.
Reference Information:
ARP 3D22 (pg. 1)
Plant Procedures 3D22 NUREG 1123 KA Catalog Rev. 2 262002 K1. Knowledge of the physical connections and/or cause-effect relationships between UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) and the following:
262002 K1.03 Rod position information: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Low New RO 2017 ILO Written Page: 95 of 207 Question 47
48 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK The plant is operating at 100% power when the following alarms are received:
- 10D68, Div II ESS 130 V Battery 2PB Trouble
- 2D24, Div II CS Logic Power Failure
- 2D50 HPCI Logic Bus Power Failure Under these conditions, control power for Safety Relief Valves (1) is lost.
Safety Relief Valve (2) remains powered from a redundant DC power source.
A. (1) A, E, H, J, P, and R (2) B B. (1) A, B, H, J, P, and R (2) E C. (1) D, F, G, K, L, M, and N (2) C D. (1) C, D, F, K, L, M, and N (2) G Answer: D 2017 ILO Written Page: 96 of 207 Question 48
Answer Explanation:
SRVs C, D, F, K, L, M, and N receive control power from Div II DC power. (I-2095-01) SRV G has a redundant control power feature 125V DC from Dedicated Shutdown PNL H21P623 (BOP) (I-2095-04), .
Distractor Explanation:
A & B are incorrect and plausible; SRVs A, B, E, H, J, P, and R are unaffected by a Div II DC power loss; they receive Div I DC control power.
C is incorrect and plausible; SRV C has no redundant DC power source, and is LOST by a Div II DC power loss.
Reference Information:
I-2095-01 & I-2095-04 Plant Procedures 20.300.260VESF 23.201 NUREG 1123 KA Catalog Rev. 2 263000 K2. Knowledge of electrical power supplies to the following:
263000 K2.01 Major D.C. loads 295004 AK1. Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER :
295004 AK1.02 Redundant D.C. power supplies: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
NRC Exam Usage ILO 2017 Exam LOR 2009 Exam LOR 2015 Exam I-Drawings I02095-4 7 5 ILO 2017 NRC Question Use:
Bank Low RO 2017 ILO Written Page: 97 of 207 Question 48
49 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED The plant is operating at 100% power. EDG 11 is running paralleled to Bus 64B for a surveillance run.
The following events then occur:
- 9D23, D1 4160V ESS Bus 64B Breaker Tripped, alarms.
- 64B-Pos.B6 trips open.
- 64B-Pos.B8 trips open.
- Bus 64B power ON light goes OFF.
Based on these indications which of the following is true?
A. West CRD Pump will trip if running.
B. North RRMG Set scoop tube positioner is locked.
C. A half scram is received on Division 1 if RPS A is on normal supply.
D. CRS should enter 20.307.01, Emergency Diesel Generator Failure ONLY.
Answer: C 2017 ILO Written Page: 98 of 207 Question 49
Answer Explanation:
Per 20.300.64B, Condtion D address the RPS loss if RPS A is on normal.
Distractor Explanation:
A. Is incorrect and plausible because West CRD is powered from Div 2.
B. Is incorrect and plausible because occurs from a Loss of 64C. (20.300.64C Note 8)
D. Is incorrect and plausible because not appropriate as bus has tripped; not an EDG issue, 64C AOP will directly entry in to 20.307.01 Reference Information:
20.300.64B Condition D Plant Procedures 20.300.64B NUREG 1123 KA Catalog Rev. 2 264000 K3. Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following:
264000 K3.03 Major loads powered from electrical buses fed by the emergency generator(s) 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High Modified RO 2017 ILO Written Page: 99 of 207 Question 49
50 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK Following a Loss of Coolant Accident with an electrical plant malfunction, plant conditions are as follows:
- 345 kV Mat Power Indicating Lights are OFF.
- EDG 13 is LOADED carrying Bus 65E.
- Bus 65F and Bus 14ED Power Indicating Lights are OFF.
- Bus 72F and Bus 72ED Power Indicating Lights are OFF.
- 65F-F6 Breaker is TRIPPED.
- 65F-F8 Breaker is CLOSED.
Which one of the following lists the electrical procedure which should be executed to provide MAXIMUM Low Pressure ECCS Injection to support 29.100.01, Sheet 1, RPV Control actions?
A. 20.300.65F, Loss of Bus 65F, due to Bus 65F being LOCKED OUT.
B. 20.300.72F, Loss of Bus 72F, due to Bus 72F being LOCKED OUT.
C. 20.307.01, Emergency Diesel Generator Failure, due to EDG 14 failing to start.
D. 20.300.SBO, Loss of Offsite and Onsite Power, due to a combination of electrical malfunctions.
Answer: C 2017 ILO Written Page: 100 of 207 Question 50
Answer Explanation:
A loss of 345kv power indicated in loss of ESF Div 2 AC Power. An EDG 14 START FAILURE has resulted in BUS 65F being deenergized Justification:
A. is incorrect because a 65F LOCKOUT condition is excluded by 65F-F8 breaker being CLOSED.
B. is incorrect because a 72F LOCKOUT condition is excluded by 65F-F6 breaker being tripped.
D. is incorrect because a Station Blackout is excluded by DIV 1 ESF Buses energized and EDG 13 LOADED.
Reference:
20.307.01, step H1 Plant Procedures 20.307.01 - EDG Failure NUREG 1123 KA Catalog Rev. 2 262001 K1. Knowledge of the physical connections and/or cause effect relationships between A.C. ELECTRICAL DISTRIBUTION and the following:
262001 K1.01 Emergency generators (diesel/jet) 264000 Emergency Generators (Diesel/Jet)
G2.4.11 Knowledge of abnormal condition procedures.
G2.4.8 Knowledge of how abnormal operating procedures are used with EOPs 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank High RO 2017 ILO Written Page: 101 of 207 Question 50
51 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK The plant is experiencing a loss of air event. Station air header is 73 psig with all Station Air Compressors running.
At this time, air supply to PCMS isolation valves is ____________________________________.
A. provided by Station Air B. provided by Instrument Air C. provided by Non-Interruptible Air D. isolated due to low Interruptible Air System pressure Answer: C 2017 ILO Written Page: 102 of 207 Question 51
Answer Explanation:
Enclosures in 20.129.01 show PCMS is an NIAS load. At this air pressure, all air supplies are isolated except NIAS.
Distractor Explanation:
A. Is plausible and incorrect because station air is an air system that is not a source with the given condition.
B. Is plausible and incorrect because Instrument Air is an air system that is not a source with the given condition.
D. Is plausible and incorrect because Non-Interruptible Air supplies, however not with the given condition.
Reference Information:
20.129.01 (Condition A-C for what happens)
M-2015 (FOS)
Plant Procedures 20.129.01 23.129 NUREG 1123 KA Catalog Rev. 2 300000 K4. Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following:
300000 K4.02 Cross-over to other air systems 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Bank Low RO 2017 ILO Written Page: 103 of 207 Question 51
52 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: NEW The plant is operating at 100% power with the following Station Air Compressor lineup:
East Station Air Compressor Danger Tagged Out (Unavailable)
Center Station Air Compressor Running West Station Air Compressor Auto Following a seismic event:
Bus 72R has been de-energized The West Station Air Compressor is damaged and will not operate.
Based on this, which of the following indicates how the Compressed Air System responds?
P50-R802 P5002-D001 STATION AIR DIV 1 CONTROL AIR PRESSURE COMPRESSOR A. 95 psi and steady RUNNING B. 84 psi and lowering RUNNING C. 84 psi and lowering STOPPED D. 95 psi and steady STOPPED Answer: B 2017 ILO Written Page: 104 of 207 Question 52
Answer Explanation:
Based on logic prints listed & AOP 20.129.01, Control Air Compressors start at 85 psi therefore system press has to go to below 85 psi and then, per the questions stem pressure recovers. Based on this the 401 will get a closed signal, and the 402 and 440 will not get close signals.
Distracter Explanation:
Distracter are plausible based on not understanding setpoints or power supplies/auto-starts of compressors A. Is incorrect because 401 will get close signal at 85#. This answer is plausible if the examinee incorrectly assumes the system pressure did not go low enough to cause isolations.
C. Is incorrect because 402 will not get close signal until 75# and air header only went to 83#. This answer is plausible if the examinee incorrectly assumes the system pressure did go low enough to cause isolations.
D. Is incorrect because 402 & 440 will not get close signal until 75# and air header only went to 83#.
This answer is plausible if the examinee incorrectly assumes the system pressure did go low enough to cause isolations.
Reference Information:
I-2450-02,04,05 are not shown here, however they are included in development folder, the AOP actions match the actuations.
20.129.01 (pg 3) System actuation based on pressure per action statement.
Plant Procedures 20.129.01 20.300.72A NUREG 1123 KA Catalog Rev. 2 295019 AA1. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:
295019 AA1.03 Instrument air compressor power supplies 300000 K5. Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM:
300000 K5.01 Air compressors 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Low New RO 2017 ILO Written Page: 105 of 207 Question 52
53 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK With the plant operating at 100% power and two TBCCW Pumps in service, the following occurs:
- 480 VAC Bus 72N is de-energized.
- 5D5, TBCCW PUMPS DIFFERENTIAL PRESSURE HIGH/LOW, alarms.
- TBCCW Headers Pressure Indicator on P43-R805 indicates 17 psid.
- TBCCW Pump Suction Pressure is 14 psig.
As a result of this transient, which of the following describes the status of the TBCCW system?
A. One TBCCW pump is in service; P43-F405, TBCCW DP Control Valve, is SHUT.
B. One TBCCW pump is in service; P43-F405, TBCCW DP Control Valve, is OPEN.
C. The remaining TBCCW pump tripped on Low Suction Pressure; P43-F405, TBCCW DP Control Valve, is SHUT.
D. The standby TBCCW pump can be started; P43-F405, TBCCW DP Control Valve, is controlling pressure in the normal band.
Answer: A 2017 ILO Written Page: 106 of 207 Question 53
Answer Explanation:
With power loss to Bus 72N, two of three TBCCW Pumps are deenergized.
Distractor Explanation:
B. Is plausible and incorrect because it would be true if Pressure were >30 psid.
C. Is plausible and incorrect because it would be true if Suction Pressure were < 7 psig.
D. Is plausible and incorrect because it would be true if Bus 72 M was deenergized and 72 N were energized Reference Information:
20.300.72N Attachment 1, Page 1 of 1 (TBCCW PUMP LOAD INFO)
ARP 5D5 (System response)
Plant Procedures 04D005 NUREG 1123 KA Catalog Rev. 2 400000 K6. Knowledge of the effect that a loss or malfunction of the following will have on the CCWS:
400000 K6.04 Pumps 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Bank High RO 2017 ILO Written Page: 107 of 207 Question 53
54 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK The shift operating crew is performing 24.106.01, Operable Control Rod Check. The P603 operator stops exercising control rods and reports the following:
- Cooling water flow is changing whenever a control rod is moved.
- Drive water flow is 2 gpm during withdraw movement.
- Drive water flow is 4 gpm during insert movement.
- CRD system flow is indicating 63 gpm and oscillating when the control rod is moving.
Which ONE of the following conditions would cause the cooling water flow oscillations?
A. Pressure equalizing valve failure.
B. Stabilizing valves failure to cycle.
C. Fluctuating air supply pressure to the CRD Flow Control Valve.
D. C11-F127 (Scram Outlet Valve) valve leaking causing increased cooling water flow.
Answer: B 2017 ILO Written Page: 108 of 207 Question 54
Answer Explanation:
The flow indications show the stabilizing valves are not operating.
Distractor Explanation:
The examinee must understand system flow paths and what the normal values are and how the system operates to answer correctly. The distractors are plausible because they are things that can be wrong with the system.
Reference Information:
C11-00 DBD for CONTROL ROD DRIVE HYDRAULICS SYSTEM (pg 103 of pdf or Page 4-77 of document) details the COMPONENT DESIGN BASIS DATA for Solenoid Valve: Control Rod Drive Hydraulic Stabilizing Valve.
Maintenance Plant Drawings 6M721-5703 CONTROL ROD DRIVE SYSTEM FUNCTIONAL OPERATING SKETCH M-2081 - Control rod Drive Hydraulic System Reactor Building Part 1 of 2 NUREG 1123 KA Catalog Rev. 2 201001 A3. Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including:
201001 A3.01 Valve operation 201001 K4. Knowledge of CONTROL ROD DRIVE HYDRAULIC SYSTEM design feature(s) and/or interlocks which provide for the following: VOLTAGE AND ELECTRIC GRID DISTURBANCES:
201001 K4.02 Stable system flow when moving control rods (stabilizing valves).
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (6) Design, components, and function of reactivity control mechanisms and instrumentation.
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Bank Low RO 2017 ILO Written Page: 109 of 207 Question 54
55 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED The operating crew is currently performing a Rod Pattern Adjustment from Group 10 to Group 9 Control Rods. While withdrawing a Control Rod, 3D80 Control Rod Drift Alarms. On the full core display a non-selected rod indicates Control Rod Drift. When the drifting rod is selected, RPIS show the rod is moving INTO the core.
What impact does this have on the Reactor Manual Control System, and what actions shall be taken?
A. A rod block is enforced. Disarm the Control Rod at its current position.
B. The ability to move control rods remains functional. Fully insert the Control Rod using RONOR Switch in EMER ROD IN.
C. Only the RONOR switch can be used to move control rods. Individually scram the Control Rod.
D. Only the "Emergency In" position of the RONOR switch is functional. Place the Reactor Mode switch in SHUTDOWN.
Answer: B 2017 ILO Written Page: 110 of 207 Question 55
Answer Explanation:
Only a Control Rod Drift occurred. The RMCS remains functional, and the Control Rod is required to be fully inserted per IMMEDIATE ACTIONS of 20.106.07.
Distracter Explanation:
A. is incorrect because a Rod Block alarm is not received.
C. is incorrect because although the RONOR switch is functional the RMCS is also functional, the rod is not scrammed unless it drifts after it has been disarmed.
D. is incorrect because the RMCS is functional and the mode switch is only placed in SHUTDOWN if 2 or more rods are drifting.
Reference Information:
20.106.07 IA.
Plant Procedures 20.106.07 NUREG 1123 KA Catalog Rev. 2 201002 A2. Ability to (a) predict the impacts of the following on the REACTOR MANUAL CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
201002 A2.02 Rod drift alarm 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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High Modified RO 2017 ILO Written Page: 111 of 207 Question 55
56 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED During a plant event, both divisions of EECW initiate:
Due to malfunction P4400-F606A, EECW DIV1 DW SPLY OUTBOARD ISO MOV closes.
Which of the following describes the impact of this fault and any required operator actions?
A. RR MG Set bearing damage could occur. Trip RR Pump A.
B. RR Pump motor damage could occur. If flow cannot be restored within 2 minutes trip RR Pump A.
C. RR Pump seal damage could occur. If seal cavity temperature exceeds 180°F, Recirc Pump A must be tripped and isolated.
D. RR MG Set North Cooling Fan air temperatures could increase. If MG set air temperature exceeds 260°F, trip RR Pump A.
Answer: B 2017 ILO Written Page: 112 of 207 Question 56
Answer Explanation:
Without cooling, RR MG set should be tripped within 2 minutes. Motor damage may occur. EECW load when EECW in service.
Distractor Explanation:
A. Is plausible, but cooling would still be available to MG set bearings RBCCW load when EECW in service.
C. Is plausible, Seal Temperature alarms at 160°F and pump should be tripped and isolated at >200°F.
D. Is plausible, but cooling would still be available to MG set cooling unit RBCCW load when EECW in service.
Reference Information:
3D122 RECIRC PUMP A SEAL CLG H2O FLOW LOW (INITIAL RESPONSE)
Plant Procedures 03D122 20.127.01 23.138.01 NUREG 1123 KA Catalog Rev. 2 202001 K5. Knowledge of the operational implications of the following concepts as they apply to RECIRCULATION SYSTEM :
202001 K5.03 Pump/motor cooling: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (4) Secondary coolant and auxiliary systems that affect the facility.
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High Modified RO 2017 ILO Written Page: 113 of 207 Question 56
57 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: MODIFIED The plant is operating at 100% power when the following alarms are received:
- 3D156, REACTOR WATER LEVEL LOW
- 3D127, RECIRC SYS LOOP B ONLY OPERATING What is the expected response from the Reactor Recirculation system and the basis for that response?
A. Both RR pumps trip to prevent inadvertent power oscillations.
B. RR Pumps run back to Limiter 1 to prevent insufficient subcooling/NPSH loss.
C. RR Pumps run back to Limiter 2 to reduce steam flow and water loss from the vessel.
D. RR Pump run back to Limiter 4 to prevent vibrations in the reactor vessel internals.
Answer: D 2017 ILO Written Page: 114 of 207 Question 57
Answer Explanation:
RR Limiter 4 is actuated by the North(South) RR MG Set drive motor or MG Set field breakers opening.
Annunciator 3D127 indicates the North RR pump has tripped. The basis for Limiter 4 is to prevent excessive vibrations on the vessel internals.
Distractor Explanation:
A Is plausible, Alternate Rod Insertion (ARI) trips both RR MG sets for RPV Level 2.
B Is plausible, RR Limiter 1 is actuated for feed flow less than 20% or Recirculation pump discharge valve not full open.
C Is plausible, RR Limiter 2 is actuated when RPV Level 4 (192.7) alarms and a RFP is tripped or has low suction flow. A runback of Recirculation Pump speed reduces steam flow thus water loss from the vessel.
Reference Information:
23.138.01 Pg 4-9 Plant Procedures 20.107.01 20.138.01 23.138.01 NUREG 1123 KA Catalog Rev. 2 295009 AK2. Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following:
295009 AK2.03 Recirculation system NRC Exam Usage ILO 2017 Exam ILO 2017 NRC Question Use:
High Modified RO 2017 ILO Written Page: 115 of 207 Question 57
58 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is operating at 97% power. RWCU is operating normally with G3352-F044, RWCU F/D Sys Bypass Valve open to support removing Filter Demins from service.
The RBCCW System is then lost.
Which of the following statements describes the expected final condition of the RWCU System, assuming NO operator actions are taken?
A. G3352-F119, RWCU Supply Suct Iso Valve closed. RWCU Pumps tripped.
B. RWCU Pumps tripped. No valve closures occur.
C. RWCU Pumps in operation. RWCU Filter Demins in HOLD.
D. G3352-F001, RWCU Supply Inbd Iso Valve, G3352-F004, RWCU Supply Otbd Iso Valve, and G3352-F220, RWCU To FW Otbd Cntm Iso Valve closed.
RWCU Pumps tripped.
Answer: A 2017 ILO Written Page: 116 of 207 Question 58
Answer Explanation:
Loss of RBCCW cooling to NRHXR causes a high temperature which leads to this effect.
When RWCU NRHX Outlet Temp reaches 140°F, G3352-F119 closes, RWCU Pumps trip, and RWCU Demins shift into Hold.
Distractor Explanation:
All distractors are plausible and are based on the examinee's ability to properly interpret indications and understand system operations. Distractor RWCU componest listed all have auto trips or events.
Reference Information:
2D110 RWCU NON REGEN HX OUTLET TEMP HIGH (Auto Actions)
Plant Procedures 02D110 20.127.01 NUREG 1123 KA Catalog Rev. 2 204000 K6. Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER CLEANUP SYSTEM :
204000 K6.01 Component cooling water systems 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Bank High RO 2017 ILO Written Page: 117 of 207 Question 58
59 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: BANK The plant is operating in Mode 3 with Division 1 RHR in Shutdown Cooling mode when RPV level begins lowering due to unknown reasons. The following indications are observed:
- 1. Narrow range RPV level show a 6 difference between instruments
- 2. Wide range shows a 20 difference between indicators
- 3. Occasional notching is occurring on all instruments Which of the following operator actions shall be taken at this time?
A. Insert all control rods not fully inserted.
B. Verify Main Turbine Bypass Valves are closed.
C. Verify Reference Leg Backfill is in service with proper flow rate.
D. Isolate SDC by closing Shutdown Cooling Suction Isolation Valves E1150-F008 and E1150-F009.
Answer: D 2017 ILO Written Page: 118 of 207 Question 59
Answer Explanation:
22.000.02 P&L 3.4.1 states "If level anomalies (level indication deviates more than 5" on narrow range indication, 22" on wide range indication or notching occurs) and an unexplained loss of RPV level occurs, take immediate action to isolate SDC by closing E1150-F008, RHR SDC Otbd Suction Iso Vlv, and E1150-F009, RHR SDC Inbd Suction Iso Vlv, and refer to 20.205.01, "Loss of Shutdown Cooling."
Distractor Explanation:
A. plausible if thought is to ensure reactor is fully shut down.
B. plausible if thought is that failed open TBVs are causing RPV level loss.
C. plausible due to malfunction of backfill causing loss of level indications.
Reference Information:
22.000.02 P&L 3.4.1 Plant Procedures 22.000.02 NUREG 1123 KA Catalog Rev. 2 216000 A4. Ability to manually operate and/or monitor in thecontrol room:
216000 A4.01 Recorders 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank Low RO Selected for Early Review (NRC) 2017 ILO Written Page: 119 of 207 Question 59
60 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW In addition to providing a positive margin to the negative design pressure of the primary containment, why is Torus Sprays terminated prior to 0 PSIG?
A. To prevent bypassing the suppression pool via the vacuum relief system in the event of a steam blowdown from the drywell.
B. To maintain a positive suppression chamber to preclude damage or failure of SRVs from "water slugs" when SRVs are opened to EMERGENCY DEPRESSURIZE.
C. To prevent bypassing the suppression pool via the vacuum relief system should it become necessary to vent the drywell so that some filtering of fission products occurs.
D. To maintain a positive suppression chamber and drywell pressure to preclude air from being drawn in through the vacuum relief system de-inerting the primary containment.
Answer: D 2017 ILO Written Page: 120 of 207 Question 60
Answer Explanation:
Per the EPGs the purpose of the override is to maintaining a positive suppression chamber and drywell pressure to precludes air from being drawn in through the vacuum relief system to de-inert the primary containment and to maintain positive margin to the negative design pressure of the primary containment.
Distractor Explanation:
A. Is incorrect and plausible because vacuum relief system does provide a flowpath, however this flowpath is closed or would close in the listed conditions.
B. Is incorrect and plausible because water slugs are common reasons for vacuum breakers, however this is not the purpose of these vacuum breakers C. Is incorrect and plausible because vacuum relief system does provide a flowpath, however this flowpath is closed or would close in the listed conditions.
Reference Information:
EPG B-7-5 Plant Procedures BWROG EPG App B NUREG 1123 KA Catalog Rev. 2 223001 K5. Knowledge of the operational implications of the following concepts as they apply to PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES :
223001 K5.01 Vacuum breaker/relief operation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Low New RO 2017 ILO Written Page: 121 of 207 Question 60
61 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW The plant is in MODE 5 with the gates installed between the Reactor Cavity and the Spent Fuel Pool.
RHR B pump is in service for Shutdown cooling.
An object is being removed from the Spent Fuel Pool. (1) Which of the following could cause fuel storage pool water level to lower and (2) What is the basis for maintaining fuel storage pool water level > 22 ft over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks?
A. (1) Trip of the Fuel Pooling Cooling and Cleanup system.
(2) Shielding gamma that is produced from decay in an irradiated fuel assembly, while the assembly is being moved.
B. (1) Trip of the Fuel Pooling Cooling and Cleanup system.
(2) Absorbing iodine gases before they are released to the secondary containment due to a fuel handling accident.
C. (1) Trip of RHR B pump.
(2) Shielding gamma that is produced from decay in an irradiated fuel assembly, while the assembly is being moved.
D. (1) Trip of RHR B pump.
(2) Absorbing iodine gases before they are released to the secondary containment due to a fuel handling accident.
Answer: B 2017 ILO Written Page: 122 of 207 Question 61
Answer Explanation:
MOP 13 gives guidance on this type of evolution. Anything that could change Spent Fuel Pool level or RWL has to be reported to the Shift Manager before the evolution. Operations assigns a monitor from the operations department, normally a NLO.
The bases state that this is based on absorbing iodine gases before they are released to the secondary containment due to a fuel handling accident.
Distracter Explanation:
(1) Trip of RHR B pump, is plausible and incorrect because it is in service, however with the gates installed shutdown cooling will not effect SFP level (2) Shielding gamma that is produced from decay in an irradiated fuel assembly, while the assembly is being moved, is plausible and incorrect because gamma is present, however this is not the concern addressed in the basis.
Reference Information:
For this question the T.S.B. is used as a reference for system information that is taught in the systems course under lesson plan LP-315-0115 per training objective C013:
C013. Describe the Fuel Pool Cooling and Cleanup system technical specification limiting conditions for operation, their bases, the associated surveillance requirement(s), and their relationship to operability.
T.S. 3.7.7 and BASES MOP 13 NUREG 1123 KA Catalog Rev. 2 233000 Fuel Pool Cooling and Cleanup G2.1.27 Knowledge of system purpose and or function.
Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Low New RO 2017 ILO Written Page: 123 of 207 Question 61
62 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK The plant is starting up after a refueling outage. The CRLNO is preparing for Main Turbine runup with the following plant conditions:
- Mode Switch in RUN
- Reactor Power ~ 10%
- Bypass Valves > 25% open IF the CRLNO places the governor interlock switch in EXERCISE PERMIT at this time, which of the following results is to be expected?
A. The VALVE OPEN LIMIT IN CONTROL and WRSG IN CONTROL backlights will illuminate.
B. N3000-F412, Hood Spray Sec Control Vlv, will open, and 4D31, LP HOOD SPRAYS ON, will alarm.
C. The in-service pressure regulators' setpoint will be raised, causing the Bypass Valves to open.
D. The in-service pressure regulators' setpoint will be lowered, causing the Bypass Valves to open.
Answer: D 2017 ILO Written Page: 124 of 207 Question 62
Answer Explanation:
Per 23.109 DO NOT place the Governor Interlock switch in EXERCISE PERMIT with any pressure in the 52" Manifold. This will cause the regulators' setpoint to be lowered approximately 30 psig and the Bypass Valve to fully open. An RPV water level and pressure transient will result. Question is based on OPEX from RF12 Distractor Explanation:
A. Is plausible and incorrect because VALVE OPEN LIMIT IN CONTROL and WRSG IN CONTROL backlights are related to the turbine and are not related to exercise permit.
B. Is plausible and incorrect because hoods spay is related to the turbine and the listed alarm is valid for the N3000-F412, however the N3000-F412 is not effected by exercise permit.
C. Is plausible and incorrect because the setpoint IS effected, however this distractor has it going the wrong way.
Reference Information:
23.109 P&L 3.21 Plant Procedures 23.109 NUREG 1123 KA Catalog Rev. 2 241000 A1. Ability to predict and/or monitor changes in parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including:
241000 A1.07 Bypass valve position 241000 A4. Ability to manually operate and/or monitor in the control room:
241000 A4.06 Bypass valves (operation) 241000 K1. Knowledge of the physical connections and/or cause effect relationships between REACTOR/TURBINE PRESSURE REGULATING SYSTEM and the following:
241000 K1.06 Bypass valves 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank Low RO 2017 ILO Written Page: 125 of 207 Question 62
63 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: BANK The plant is operating at 100% power when an electrical fault trips the NORMAL feed to 65D.
What is the short term impact on the Condensate/Feedwater delivery system?
A. Loss of cooling to the Reactor Feedwater pump bearings.
B. Trip of Heater Feed Pumps and Reactor Feedwater Pumps.
C. Trip of Heater Drains Pumps and Reactor Recirc runback to Limiter 1.
D. Reduction in Feedwater temperature and subsequent Reactor power spike.
Answer: B 2017 ILO Written Page: 126 of 207 Question 63
Answer Explanation:
The loss of 65D will cause a trip of the North and South Condenser pumps and loss of feedwater to the vessel. This will require entry into 20.107.01 Loss of Feedwater or Feedwater Control, and the immediate action of placing the mode switch in shutdown.
Distractor Explanation:
A/C/D are incorrect because a loss of condensate will result in a RFP trip and SCRAM condition.
Reference Information:
20.300.65D Attachment 1 Plant Procedures 20.107.01 20.300.65D NUREG 1123 KA Catalog Rev. 2 256000 K2. Knowledge of electrical power supplies to the following:
256000 K2.01 System pumps 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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Bank Low RO 2017 ILO Written Page: 127 of 207 Question 63
64 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK A reactor startup is in progress and RPV pressure is 200 psig. The following alarms actuate indicating fuel damage and radiation release:
- 3D20, 2 MINUTE HOLDUP PIPE RADN MONITOR UPSCALE
- 3D24, 2 MINUTE HOLDUP PIPE RADN MONITOR UPSCALE TRIP The CRS announces entry into the Secondary Containment Control/Radioactive Release EOP.
Condenser vacuum then begins to DECREASE.
What is the cause of the degrading vacuum?
A. Steam Jet Air Ejectors isolated.
B. Offgas Ring Water Pumps tripped.
C. Mechanical Vacuum Pumps tripped.
D. Main Steam Isolation Valves closed.
Answer: C 2017 ILO Written Page: 128 of 207 Question 64
Answer Explanation:
MVP trip is indicated by 3D24, 2 MINUTE HOLDUP PIPE RADN MONITOR UPSCALE TRIP, alarms. MS rad monitors have net yet tripped.
Distractor Explaination:
A. Is incorrect and plausible because SJAE are not in service until 300 psig.
B. Is incorrect and plausible can effect vacuum, however this will not cause Alarm 3D24 D. Is incorrect and plausible can effect vacuum, however this will not cause Alarm 3D24 Reference Information:
3D24 Plant Procedures 20.000.07 NUREG 1123 KA Catalog Rev. 2 271000 A1. Ability to predict and/or monitor changes in parameters associated with operating the OFFGAS SYSTEM controls including:
271000 A1.12 Process radiation monitoring indications 271000 K1.02 Process radiation monitoring system 272000 A3. Ability to monitor automatic operations of the RADIATION MONITORING SYSTEM including:
272000 A3.02 Offgas system isolation indications 272000 K1. Knowledge of the physical connections and/or cause effect relationships between RADIATION MONITORING SYSTEM and the following:
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Bank High RO 2017 ILO Written Page: 129 of 207 Question 64
65 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: MODIFIED Division I CCHVAC is operating in Purge mode due to a fire in the Relay Room when the Main Control Room receives the following alarm:
3D36 DIV I/II RB VENT EXH RADN MONITOR UPSCALE TRIP What will happen to the CCHVAC configuration?
A. Div I CCHVAC will trip and will have to be manually started in the Recirculation mode.
B. Div I CCHVAC will transfer from Purge mode to Recirculation mode.
C. Div I CCHVAC will trip and Div II CCHVAC will automatically start in the Recirculation mode.
D. Div I CCHVAC will continue to operate in Purge mode. Logic must be reset to allow shift to Recirculation mode.
Answer: B 2017 ILO Written Page: 130 of 207 Question 65
Answer Explanation:
Per Design Base Document, Hierarchy of Modes, Automatic Recirc mode will initiate from any other mode. T-41-02 Section 2.2.4.3 Recirculation Mode states recirc mode will initiate due to Reactor building ventilation exhaust radiation high.
Distracter Explanation:
Distracters are incorrect and plausible based on all distracter list actual modes of CCHVAC and describe events/modes for CCHVAC, the Hierarchy of Modes directs that an automatic shift to recirc will occur.
Reference Information T-41-02 pg. 4-5, 20 Plant Procedures 23.413 T41-02 CONTROL CENTER HEATING, VENTILATING, AND AIR-CONDITIONING (CCHVAC)
SYSTEM NUREG 1123 KA Catalog Rev. 2 290003 K4. Knowledge of CONTROL ROOM HVAC design feature(s) and/or interlocks which provide for the following:
290003 K4.01 System initiations/reconfiguration: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High Modified RO 2017 ILO Written Page: 131 of 207 Question 65
66 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK In accordance with MOP09, "Locked Valve", a ____________________ padlock should be used on a locked "throttled" position valve.
A. red B. black C. red with black dot D. yellow with black dot Answer: C 2017 ILO Written Page: 132 of 207 Question 66
Answer Explanation:
Per MOP 09 Section 3.8.3 Red with black dot - THROTTLED Distractor Explanation:
Distractor colors are incorrect and used for other configuration controls.
Reference Information:
MOP09 (pg 4)
Plant Procedures MOP09 - Locked Valve Guidelines NUREG 1123 KA Catalog Rev. 2 G2.1.1 Knowledge of conduct of operations requirements NRC Exam Usage ILO 2017 Exam ILO 2017 NRC Question Use:
Bank Low RO 2017 ILO Written Page: 133 of 207 Question 66
67 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: BANK There is a small leak in the Steam Tunnel. Using a camera, an isolation valve has been identified.
During the pre-job brief, this activity is identified as MODERATE. RP has determined that plastic PCs are required for entry. Dry Bulb temperature is 104°F, and no other heat stress prevention equipment will be used.
After isolating the leak, the Nuclear Operator reports the job took 10 minutes.
The EARLIEST this Nuclear Operator can be sent back to normal duties is after _____ minutes of rest time.
A. 10 B. 20 C. 30 D. 60 Answer: C 2017 ILO Written Page: 134 of 207 Question 67
Answer Explanation:
Using TABLE 21-3, stay time is 20 minutes for plastic PCs for a temperature of 104°F. Using the rest time calculation on page 21.5, rest time is actual time/time limit (10/20) times 60 minutes, or 30 minutes.
Distractor Explanation:
A. is incorrect, but could be correct if chart read wrong or calculation not properly utilized.
B. is incorrect because this is the actual stay time allowed.
D. is incorrect because if this does not take the ratio of time worked vs stay time into the calculation.
Reference Information:
Safety Handbook, Section 21, Table 21-3B Plant Procedures Safety Handbook, Section 21, Hot and Cold Environments / Temperature Extremes NUREG 1123 KA Catalog Rev. 2 G2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
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Bank Low RO 2017 ILO Written Page: 135 of 207 Question 67
68 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK I & C technicians are preparing to perform a surveillance procedure on an instrument. They observe that a prerequisite to the procedure cannot be met under present plant conditions.
Omitting the prerequisite will not affect the performance of the procedure, and the surveillance will be past its critical date the next day.
Which ONE of the following selections indicates (1) who can waive the prerequisite and (2) how this waiver must be documented after the prerequisite is marked as N/A in order to continue with the surveillance?
A. (1) Control Room Supervisor (2) Make an entry in the Unit Log B. (1) Field Support Supervisor (2) Make an entry in the controlling surveillance document C. (1) I&C Supervisor (2) Write a CARD to document and investigate the waiver D. (1) Shift Technical Advisor (2) Document the waiver on the Surveillance Performance Form (SPF)
Answer: A 2017 ILO Written Page: 136 of 207 Question 68
Answer Explanation:
MGA03, step 5.1.7, states the Shift Manager/Control Room Supervisor can waive the prerequisite if warranted. The reason for waiving the prerequisite shall be justified and documented either in the controlling document or Unit Log.
Distractor Explanation:
B. Is incorrect because the FSS cannot waiver a prereq, but is plausible because the waiver can be documented on the controlling document.
C. Is incorrect because a CARD is not used to document the waiver per MGA03, but is plausible because the I&C Supervisor is the technician's supervisor.
D. Is incorrect because the waiver cannot be documented on the SPF per MGA03, but is plausible because the Shift Manager can waive a prerequisite.
Reference Information:
MAG03 (5.1.7)
Plant Procedures MGA03 - Procedure Use and Adherence NUREG 1123 KA Catalog Rev. 2 G2.2.12 Knowledge of surveillance procedures 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank Low RO 2017 ILO Written Page: 137 of 207 Question 68
69 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK The plant has been shut down for an extended outage. Prior to work being released, which of the following systems must be verified free of hydrogen using a gas sampler?
A. TBCCW B. Main Turbine C. Heater Drains D. Main Steam Lines Answer: C 2017 ILO Written Page: 138 of 207 Question 69
Answer Explanation:
Per MOP12, the following systems have a potential for the presence of hydrogen in the piping and components: Off Gas, Feed Water, Condensate downstream of the CFDs, Heater Drains, H2 Seal Oil, Stator Winding Cooling, Hydrogen Water Chemistry and Generator Gas systems. These systems must be verified free of hydrogen using a gas sampler prior to work being released on the system.
Distractor Explanation:
A. Is plausible because TBCCW cools the stator winding cooling which is another system that must be free of Hydrogen, however it is incorrect B. Is plausible because Main Turbine has an interface with hydrogen related systems, however is not system required to be checked.
D. Is plausible Main steam is the source of Hydrogen into the offgas system, however it is not required to be checked.
Referees Information:
MOP12 3.2.12 (pg 9)
Plant Procedures MOP12 - Tagging and Protective Barrier System NUREG 1123 KA Catalog Rev. 2 G2.2.13 Knowledge of tagging and clearance procedures.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank Low RO 2017 ILO Written Page: 139 of 207 Question 69
70 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW Per MOP05, Control of Equipment, while performing lineups, components located in a locked high radiation area are not required to be verified in correct position if:
A. Lineup is for a non-safety related system.
B. Indication is visible from the main control room.
C. The lineup is being performed on an operating system.
D. Documentation is available showing no personnel entries could affect component positioning.
Answer: D 2017 ILO Written Page: 140 of 207 Question 70
Answer Explanation:
Operations Conduct Manual MOP05, Control of Equipment, section 3.2 for performing lineups states that components located in a high radiation area, entry is not required if documentation is available showing no personnel entries that could affect component positioning or a previous lineup exists documenting correct position.
Distractor Explanation:
If an Initial Licensed candidate or Licensed Operator does not fully understand the requirements for performing lineups in a locked high radiation area (LHRA), the distractors could be considered plausible.
A. Non-Safety related equipment does not exempt an operator from performing a correct lineup in LHRA B. Component indication from the control room panel is not a valid method of performing a lineups IAW MOP05.
C. An operating component in a LHRA is not exempt from a lineup verification.
Reference Information:
MOP05 Sedition 3.2 (pg 18)
Plant Procedures MOP05 - Control Of Equipment NUREG 1123 KA Catalog Rev. 2 G2.2.14 Knowledge of the process for controlling equipment configuration or status.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Low New RO 2017 ILO Written Page: 141 of 207 Question 70
71 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: MODIFIED An accident has occurred which requires life saving measures. Radiation levels in the area of the injured person are 15,000 mrem/hr. Emergency exposure TEDE (Whole Body) limit for life saving operations has been authorized.
The MAXIMUM stay time for a rescuer under these circumstances is _____________.
A. 20 minutes B. 40 minutes C. 60 minutes D. 100 minutes Answer: D 2017 ILO Written Page: 142 of 207 Question 71
Answer Explanation:
Per EP-201-03 Table 3, the dose limit to save lives is 25 REM TEDE Whole Body. 25000/15000 = 1.66 (60 mins) = 100 mins Distractor Explanation:
A. 5000/15000 (60 mins) = 20 mins B. 10000/15000 (60 mins) = 40 mins C. 15000/15000 (60 mins) = 60 mins Reference Information:
Per EP-201-03 Table 3 (pg 6)
Plant Procedures EP-201-03 NUREG 1123 KA Catalog Rev. 2 G2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
G2.3.4 Knowledge of radiation exposure limits under normal and emergency conditions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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High Modified RO 2017 ILO Written Page: 143 of 207 Question 71
72 Points: 1.00 Difficulty: 3.00 Level of Knowledge: Low Source: BANK A licensed operator is conducting a Normal Day to Day (green trip ticket) inspection of equipment which is located in a high radiation area. In accordance with plant administrative procedures, the operator must perform the inspection by which of the following methods?
A. The evolution should be preplanned and the received dose should be maintained ALARA.
B. The operator should enter the area with a hand held monitoring device.
C. Obtain Radiation Protection verbal approval prior to the inspection D. Conducted by a visual inspection at the barrier to the area.
Answer: D 2017 ILO Written Page: 144 of 207 Question 72
Answer Explanation:
IAW MOP01, Normal, day to day inspection of high radiation areas should be made by visual inspection at the door for abnormal conditions such as steam, leaks, high temperature, or unusual noises.
Distractor Explanation:
A. Is incorrect and plausible because this necessary if it is required to enter the inspection area.
B. Is incorrect and plausible because entry is not allowed for day to day inspections.
C. Is incorrect and plausible because this action would be required per MRP06 if entry was needed.
Reference Information:
MOP01 Section 3.30 NUREG 1123 KA Catalog Rev. 2 G2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Bank Low RO 2017 ILO Written Page: 145 of 207 Question 72
73 Points: 1.00 Difficulty: 4.00 Level of Knowledge: High Source: NEW The plant is operating at 100% power when the Keylock Mode Selector Switch is rotated out of OPERATE for the following Main Steam Line Radiation Monitors:
- D11-K603A MAIN STEAM LINE A RADIATION MONITOR
- D11-K603B MAIN STEAM LINE B RADIATION MONITOR What is the Status of (1) RPS and (2) NSSS?
A. (1) Scram (2) Group 1: B2103-F016 MS LINE INBD DRAIN ISO VALVE Closed Group 2: B3100-F019 RR SAMPLE LINE INBD ISO VLV Closed B. (1) Half Scram (2) Group 1: B2103-F016 MS LINE INBD DRAIN ISO VALVE Closed Group 2: B3100-F019 RR SAMPLE LINE INBD ISO VLV Closed C. (1) Scram (2) Group 1: B2103-F019 RR SAMPLE LINE INBD ISO VLV Closed Group 2: B3100-F020 RR SAMPLE LINE OTBD ISO VLV Closed D. (1) Half Scram (2) Group 1: B2103-F019 RR SAMPLE LINE INBD ISO VLV Closed Group 2: B3100-F020 RR SAMPLE LINE OTBD ISO VLV Closed Answer: A 2017 ILO Written Page: 146 of 207 Question 73
Answer Explanation:
A full SCRAM Based on logic (A1 or A2) and (B1 or B2) would result with isolations A and B =Group 1:
B2103-F016 & Group 2: B3100-F019. And per 23.625 section 3.2 Placing the Mode Selector switch for the Main Steam Line Radiation monitors in any position other than OPERATE will actuate RPS and NSSS trip logic.
Distractor Explanation:
Distractors are plausible and incorrect based on trip sheet from 23.601. Trips isolations are for other combinations of MSLRM.
Reference Information:
23.601 (Pg 46) 23.625 (Pg 15) P&L Section 3.2 Plant Procedures 23.601 23.625 NUREG 1123 KA Catalog Rev. 2 G2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High New RO 2017 ILO Written Page: 147 of 207 Question 73
74 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant is operating at 10% power. A transient occurs which causes a reactor scram. Which one of the following indications would always indicate an Emergency Operating Procedure (EOP) entry condition?
A. 3D168 REACTOR PRESSURE HIGH B. 1D42 REAC VESSEL / PRESS RECORDER HI SPD MODE C. Division 1 and Division 2 EECW White Emergency Mode lights ON D. 3D73 TRIP ACTUATORS A1/A2 TRIPPED AND 3D74 TRIP ACUATORS B1/B2 TRIPPED Answer: B 2017 ILO Written Page: 148 of 207 Question 74
Answer Explanation:
Both setpoints for this Alarm are EOP entry conditions. 173.4" and 1093#
Distractor Explanations:
A. Is incorrect and plausible because reactor pressure high is EOP, however this alarm is 1045# (1093#
is the EOP).
C. Is incorrect and plausible because EECW does activate on high drywell pressure an EOP entry, however it also activates on low RBCCW pressure.
D . Is incorrect and plausible because at 100% this would cause an EOP entry condition on L3. However it does not at 10% power.
Reference Information:
ARP 1D42 setpoints Plant Procedures 01D42 NUREG 1123 KA Catalog Rev. 2 G2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
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High New RO 2017 ILO Written Page: 149 of 207 Question 74
75 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW Which one of the following identifies the MINIMUM required qualifications for person(s) responsible for the Command Function in the control room during (1) NORMAL and (2)
EMERGENCY conditions?
A. (1) CRS (2) CRS B. (1) CRS (2) SM C. (1) SM (2) CRS D. (1) SM (2) SM Answer: A correct answer, not B 2017 ILO Written Page: 150 of 207 Question 75
Answer Explanation:
Per MOP01 There shall be one individual with an active Senior Reactor Operator (SRO) license for Fermi 2 assigned the Command Function for and located in the Control Room at all times when the plant is in Operational Condition 1, 2, or 3. During emergency conditions this shall be the SM. During routine operations it shall be the CRS except for short periods of relief, during which the SM should be in the Control Room, but another individual possessing an active SRO license for Fermi 2 may provide relief.
Distractor Explanation:
Distractors are plausible and incorrect based on answer.
Reference Information:
MOP01 Pg 20 (Section 3.7.4)
ILO 2017 Exam Notes TASK ID 02SFGA004 OBJECTIVE ID 45397 Covered on Adminatrative Qualification Card (QP0013)
Plant Procedures MOP01 Conduct of Operations NUREG 1123 KA Catalog Rev. 2 G2.4.12 Knowledge of general operating crew responsibilities during emergency operations 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
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Low New RO 2017 ILO Written Page: 151 of 207 Question 75
76 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant was operating at 100% power when a transient is observed.
The plant has stabilized with the following indications:
- Generator Megawatt Output is 1047 Mwe.
- Reactor Power is 91%.
- RPV Level is 197 inches.
- Reactor Pressure is 1017 psig.
- Total core flow 84 Mlbm/hr Based on these conditions, (1) What has occurred and (2) which of the following action must the CRS Direct?
A. (1) Jet Pump Failure (2) Adjust RRMG set speeds to match loop jet pump flows and stay outside the SCRAM and Exit Regions.
B. (1) Jet Pump Failure (2) Monitor core for thermal-hydraulic instability using 24.000.01 Att.34B C. (1) Uncontrolled Recirc Flow Change.
(2) Adjust RRMG set speeds to match loop jet pump flows and stay outside the SCRAM and Exit Regions.
D. (1) Uncontrolled Recirc Flow Change.
(2) Monitor core for thermal-hydraulic instability using 24.000.01 Att.34B 2017 ILO Written Page: 152 of 207 Question 76
Answer: B 2017 ILO Written Page: 153 of 207 Question 76
Answer Explanation:
The increased recirc flow and decreased jet pump flow are consistent with a jet pump failure. 20.138.02 requires 24.000.01 Att. 34B be performed to Monitor core for thermal-hydraulic instability (A.2) 10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP 20.138.02, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (jet pump failure) and then the correct mitigating action to take.
Distracter Explanation:
"Uncontrolled Recirc Flow Change" is incorrect and plausible, because loop flow indication did change.
The fact that B loop when up and jet up went down indicates a jet pump failure and not a Uncontrolled Recirc Flow Change.
"Adjust RRMG set speeds to match loop jet pump flows and stay outside the SCRAM and Exit Regions" is incorrect and plausible, because it is the an action required for Uncontrolled Recirc Flow Change and NOT an action for Jet Pump Failure.
Reference Information:
20.138.02 Symptoms and A.1 to A.7 Plant Procedures 20.138.02 NUREG 1123 KA Catalog Rev. 2 295001 AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :
295001 AA2.04 Individual jet pump flows: Not-BWR-1&2 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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High New SRO 2017 ILO Written Page: 154 of 207 Question 76
77 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK At 0800 on March 23rd, while operating in Mode 1, a Division 2 battery charger was damaged.
One hour later it was reported that the Division 2 battery terminal voltage is 124 volts.
Based on these conditions, when is the plant required to be in Mode 3?
A. March 23rd at 2000 B. March 23rd at 2100 C. March 23rd at 2200 D. March 23rd at 2300 Answer: D 2017 ILO Written Page: 155 of 207 Question 77
Answer Explanation:
In Mode 1, one inoperable battery charger must be restored in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (TS 3.8.4 Condition A). One inop battery charger and battery must be restored in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (Condition B). This overrides the initial required action. If both cannot be restored in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, we must be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 1 hr + 2 hr + 12 hr = 15 hr. 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> from initial failure (0800) is 2300.
10 CFR 55.43(b)(2) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of T.S and how to apply TS, it is not related to immediate actions or TS Safety Limits and the entry conditions are not relevant or leading to the answer. While information found above the double line is needed, correct understanding of rules Section 1 and Section 3 is needed. The answer to this questions is based on assessing plant conditions (Loss of DC) and applying requirements of TS.
A. Is incorrect and plausible because it could be selected if the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is calculated from 0800.
B. Is incorrect and plausible because it could be selected if the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is calculated from the low voltage report.
C. Is incorrect and plausible because it could be selected if the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is calculated from 0800, then the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> added.
Reference Infomation:
TS 3.8.4, Condition C NUREG 1123 KA Catalog Rev. 2 295004 Partial or Complete Loss of D.C. Power G2.2.40 Ability to apply technical specifications for a system Technical Specifications 3.8.4 DC Sources Operating 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.
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Bank High SRO 2017 ILO Written Page: 156 of 207 Question 77
78 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is operating at 85% power.
At 1355, a seismic event occurs.
At 1400, 1D66 Steam Leak Detection Ambient Temp High alarms. Steam tunnel temperature is 170°F and rising 5°F/Min. RWCU indicates isolated. RWCU Hx room temperature is 175°F and increasing 2°F/min.
As CRS, what action shall you direct FIRST, and reason for this action?
A. After a brief, place the Reactor Mode switch in SHUTDOWN. This will reduce the energy being discharged into secondary containment and prevent automatic scram on MSIV closure.
B. Immediately shut MSIVs. This will prevent a primary system from discharging into secondary containment.
C. At 1408, place the Reactor Mode Switch in SHUTDOWN. Required action due to exceeding max safe temperature, protecting plant equipment.
D. At 1420, open 5 SRVs (ADS preferred). This will reduce the energy being discharged into secondary containment (two area temperatures greater than max safe).
Answer: A 2017 ILO Written Page: 157 of 207 Question 78
Answer Explanation:
Indications give two area temperatures greater then max normal, which requires isolation of any primary system discharging into secondary. MSIVs shutting automatically scram the reactor. The mode switch is taken to shutdown prior to shutting the MSIVs.
B. is plausible; this is a valid action, but would cause an automatic scram, so the mode switch is taken to shutdown prior to this action.
C. is plausible; based on the leak rate, steam tunnel temp would be 210°F. This is the max safe temperature. EOPs direct you to place mode switch in shutdown prior to reaching this temperature. You would also get a MSIV isolation at this time.
D. is plausible; at 1408, the reactor would scram on MSIV closure. At 1420, you would get your second area temperature >max safe, requiring an ED.
Reference:
29.100.01 SH 5, SC-3 Plant Procedures 29.100.01 SH 5 NUREG 1123 KA Catalog Rev. 2 295006 SCRAM 295032 EK3. Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE :
295032 EK3.02 Reactor SCRAM G2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 158 of 207 Question 78
79 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED A fire is in progress. Turbine Building Area Temperature has exceeded 200°F and an automatic Reactor Scram has occurred. All control rods are fully inserted and the Reactor Mode Switch is in SHUTDOWN.
Conditions exist that require the evacuation of the Main Control Room to the Remote Shutdown Panel. The Relay room remains accessible. At the Remote Shutdown Panel, the following conditions are noted:
- RPV Pressure is 900 psig.
- RPV Water Level is 175 inches.
Assuming that COLD Shutdown is desired, which one of the following is (1) the procedural directed method for conducting a cooldown, and (2) What is the LOWEST RPV Pressure allowable within ONE hour?
A. (1) LOWER the pressure setting on pressure controllers in Turbine Control Relay Panel.
(2) 400 psig.
B. (1) LOWER the pressure setting on pressure controllers in Turbine Control Relay Panel.
(2) 350 psig.
C. (1) OPEN Safety Relief Valves A or B at the Remote Shutdown Panel.
(2) 400 psig.
D. (1) OPEN Safety Relief Valves A or B at the Remote Shutdown Panel.
(2) 350 psig.
Answer: C 2017 ILO Written Page: 159 of 207 Question 79
Answer Explanation:
RISING Turbine Building Area Temperatures causing a Reactor Scram indicate MSIVs are CLOSED.
SRVs will be operated. 20.000.19 limits cooldown to 90°F/hr. The LOWEST RPV Pressure allowable within ONE hour is 450 psig.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP 20.000.19, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (Cooldown during control room abandonment ) and then taking action based on the requirements of AOP 20.000.19 Distracter Explanation:
(1) LOWER the pressure setting on pressure controllers in Turbine Control Relay Panel is plausible and incorrect if the examinee thought the cause of the scram was not closed MSIVs and thought they were OPEN.
(2) 350°F is plausible and incorrect if the examinee thought the cooldown limit is 100°F/hr.
Reference Information:
AOP 20.000.19 Cond K pg 13 and Attachment 1, pg 1 Plant Procedures 20.000.19 NUREG 1123 KA Catalog Rev. 2 295016 AA2. Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT :
295016 AA2.06 Cooldown rate 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Low Modified SRO 2017 ILO Written Page: 160 of 207 Question 79
80 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is operating at 100% power.
- P50-R802, Station Air Header Pressure is 90 psig (lowering).
Which ONE of the following is required per 20.129.01, Loss of Station and/or Control Air and the reason for that action?
It is required to:
A. Start any available Station Air Compressors to prevent the INBOARD Main Steam Isolation Valves from drifting shut.
B. Start any available Station Air Compressors to prevent the OUTBOARD Main Steam Isolation Valves from drifting shut.
C. Close P5000-F401, Station Air to TB Hdr Iso VLV to prevent the INBOARD Main Steam Isolation Valves from drifting shut.
D. Close P5000-F401, Station Air to TB Hdr Iso VLV to prevent the OUTBOARD Main Steam Isolation Valves from drifting shut.
Answer: B 2017 ILO Written Page: 161 of 207 Question 80
Answer Explanation:
With Station Air Low Pressure, 20.129.01 requires starting ANY Station Air Compressor. This action prevents OUTBOARD MSIVs from drifting shut.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP 20.129.01, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (Air loss) and then taking action based on the requirements of AOP 20.129.01 and know the reasons Distractor Explanation:
A. is plausible; INBOARD MSIVs are provided by DIV1 Drywell Pneumatics.which is supported by DIV 1 NIAS.
C. is plausible. With these conditions, it is FIRST required to start ANY Station Air Compressor.
OUTBOARD MSIVs are the concern.
D. is plausible. With these conditions, it is FIRST required to start ANY Station Air Compressor.
Reference Information:
20.129.01 Cond A & Caution 1 20.129.01 BASIS Cond A & Caution 1 ILO 2017 Exam Note:
The K/A is IAS. Question appeared to be on Station Air to outside review. Question is IAS, there are not IAS air compressors. IAS is supplied from Station Air. Answers and distractors are all IAS related.
Plant Procedures 20.129.01 NUREG 1123 KA Catalog Rev. 2 295019 AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR :
295019 AA2.01 Instrument air system pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 162 of 207 Question 80
81 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED The plant was shutdown due to a leak in the drywell. The following conditions exist:
- All rods inserted.
- Both Reactor Feed Pumps are tripped, Heater Feed Pumps are available.
- RPV Pressure is 890 psig and slowly lowering.
- Drywell Pressure is 9 psig and rising.
- Drywell Temperature is 170°F and rising.
- Torus Pressure is 7.5 psig and rising.
As CRS, you shall direct which of the following?
A. Initiate Drywell Sprays.
B. Initiate Torus Cooling and Torus Sprays.
C. Isolate EECW to and from the Drywell.
D. Lower RPV pressure to allow Heater Feed Pumps to recover level.
Answer: B 2017 ILO Written Page: 163 of 207 Question 81
Answer Explanation:
Per 29.100.01 SH 2, step PCP-4, you must spray the torus prior to torus pressure reaching 9 psig. The standard order is place in torus cooling /torus spray.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of actions required by the EOPs, it is not related to immediate actions and the entry conditions are not relevant or leading to answer. The SRO must assess the plant conditions (torus pressure) and then selecting the step of the EOP to mitigate / recover.
Distractor Explanation:
A. Is incorrect and plausible because you have not exceeded the 9 psig in the Torus.
C. Is incorrect and plausible because you have not exceeded 242°F drywell temperature. You will however isolate EECW to the drywell in step PCP-3.
D. Is incorrect and plausible because with level recovering, your priority would be addressing containment. This would be a strategy if level was not recovering.
Reference Information:
29.100.01 SH 2, PCP-4 Plant Procedures 29.100.01 SH 2 NUREG 1123 KA Catalog Rev. 2 295024 EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:
295024 EA2.04 Suppression chamber pressure: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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High Modified SRO 2017 ILO Written Page: 164 of 207 Question 81
82 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant was operating normally with RCIC isolated for scheduled maintenance when a Station Blackout occurred. HPCI has failed to inject due to electrical problems with the injection valve. At time 12:00, the following conditions exist:
- Reactor pressure 950 psig (slowly rising due to decay heat)
- Reactor water level -32" (slowly lowering at 0.2"/minute)
- Torus water temperature 120°F
- Drywell temperature 160°F
At 12:20, the HPCI injection valve is opened, and water level begins RISING.
Which of the following actions is the crew directed to perform NEXT?
A. Enter the SAGs.
B. Emergency depressurize the reactor.
C. Stabilize pressure less than high pressure scram setpoint.
D. Depressurize the RPV by fully opening Mn Turb B/P valves, ignoring cooldown rate.
Answer: C 2017 ILO Written Page: 165 of 207 Question 82
Answer Explanation:
Based on conditions above, the requirements for STEAM COOLING have been met (RPV level <0 and no injection from table 1,2, or 3) When injection starts and level starts to rise the CRS will take the STC-OR1, which directs P-3 onsheet 1 - stabilize pressure less than high pressure scram setpoint (1093 #).
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of actions required by the EOPs, it is not related to immediate actions and the entry conditions are not relevant or leading to answer. The SRO must assess the plant conditions (RPV Level and Injection to the RPV) and then selecting the step of the EOP to mitigate / recover.
Distractor Explanation:
A. Is incorrect and plausible because a SAG entry would be required if you could not recover level greater than -25 inches. The stem state level is rising.
B. Is incorrect and plausible . This action would be required if level dropped to -40 inches with no injection.
D. Is incorrect and plausible . Do not meet the condition to execute P-OR1-2, it is however a valid override for the section of the EOP that is being used.
Reference Information:
29.100.01 SH1 29.100.01 SH3A Plant Procedures 29.100.01 SH 1 29.100.01 SH 3 NUREG 1123 KA Catalog Rev. 2 295031 EA2. Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL :
295031 EA2.04 Adequate core cooling 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 166 of 207 Question 82
83 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is operating at 60% power. Digital Feedwater Level Control is in 3-Element control with the Reactor Level Select Switch in A. 3D164, FEEDWATER CONTROL DCS TROUBLE has alarmed, and both Reactor Feedwater Pump Controllers have switched to Emergency Bypass.
RPV Level is 196 inches and slowly lowering.
The CRS will direct which of the following procedural actions?
A. Perform SOP 23.107. PLACE the Level Control Mode Switch in 1 ELEM, and continue power operation in Single Element Control.
B. Perform SOP 23.107. PLACE both Emergency Bypass switches in position M, manually adjust Reactor Feedwater Pump Controllers to match Feed Pump speeds, and continue power operation in 3 Element Control.
C. Enter AOP 20.107.01. Place the Reactor Mode Switch in Shutdown. Allow Post Scram Feedwater Logic to recover level.
D. Enter AOP 20.107.01. PLACE both Emergency Bypass switches in position M, manually adjust Reactor Feedwater Pump Controllers to match Feed Flow with Steam Flow, and continue power operation with manual feed pump speed control.
Answer: D 2017 ILO Written Page: 167 of 207 Question 83
Answer Explanation:
With both Reactor Feed Pump Controllers in EMERGENCY BYPASS, manual feedwater control is required by the AOP to maintain level.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP (20.107.01), it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions ( DCS TROUBLE and FEED PUMP CONTROLERS) and then selecting the correct section of AOP..
Distracter Explanation:
A. Is plausible and incorrect; would be true for a loss of one feed flow or two steam flow inputs, if NOT in Emergency Bypass.
B. Is a plausible and incorrect misconception in that matching RFP speeds would be effective vice matching steam flow and feed flow.
C. Is plausible and incorrect misconception; testing knowledge of pumping capacity with controllers in Emergency Bypass.
References Information:
20.107.01, page 9; 23.107, page 8 Plant Procedures 20.107.01 23.107 NUREG 1123 KA Catalog Rev. 2 295009 AA2. Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL :
295009 AA2.02 Steam flow/feed flow mismatch 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 168 of 207 Question 83
84 Points: 1.00 Difficulty: 4.00 Level of Knowledge: High Source: NEW The plant is operating at 50% reactor power. HPCI is in service for testing.
Review attached copy of TR 3.3.6.4 and TS 3.6.2.1.
TRSR 3.3.6.4.1, SR 3.6.2.1.1 were performed earlier in shift, before the HPCI testing.
Point 4[D] on T23-R800 (T23N004) failed its CHANNEL CHECK per TRSR 3.3.6.4.1.
Based on these surveillances and above indication of Torus temperatures, which of the following meet a CONDITION that has a REQUIRED ACTION?
A. ONLY TS 3.6.2.1 B. ONLY TR 3.3.6.4 C. BOTH TR 3.3.6.4 AND TS 3.6.2.1 D. NEITHER TR 3.3.6.4 NOR TS 3.6.2.1 Answer: C 2017 ILO Written Page: 169 of 207 Question 84
Answer Explanation:
Based on requirements of TR 3.3.6.4 Point 4 is INOP. and Point 4 must be substituted for with POINT 5.
Calculation:
Indicated Corrected Point 1 106.8 106.8 Point 2 103.6 103.6 Point 3 103.6 103.6 Point 4 98.2 108.2 Point 5 108.2 108.2 Point 6 102.9 102.9 Point 7 103.8 103.8 Point 8 105.6 105.6 Highest 108.2 108.2 Lowest 98.2 102.9 Difference 10 5.3 Average 104.1 105.3 Based on this neither TR 3.3.6.4 nor TS 3.6.2.1 are met. Therefore an ACTION statment must be entered for BOTH.
10 CFR 55.43(b)(2) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of T.S and TRM and how to apply actions required by a surveillance, it is not related to immediate actions or TS Safety Limits and the entry conditions are not relevant or leading to the answer. While information found above the double line is needed, correct understanding of an SR is required to assess LCO 3.6.2. The answer to this questions is based on assessing plant conditions (Failed instrument, elevated torus temperatures and HPCI running) and evaluating surveillance results and then applying requirements of a surveillance from the TRM.
Distracter Explanation:
Distracters are plausible and incorrect based on applying T.S and TRM Reference Information:
TR 3.3.6.4, TS 3.6.2, TRSR 3.3.6.4.1, SR 3.6.2.1.1 2017 ILO Written Page: 170 of 207 Question 84
NUREG 1123 KA Catalog Rev. 2 295013 High Suppression Pool Water Temperature G2.2.37 Ability to determine operability and/or availability of safety related equipment Technical Specifications 3.6.2.1 Suppression Pool Average Temperature Technical Requirements Manual 3.3.6.4 Suppression Pool Water Temperature Instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.
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High New SRO 2017 ILO Written Page: 171 of 207 Question 84
85 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: BANK A scram condition exists, RPS did not actuate and attempts to manually scram have failed. The crew enters the EOPs on failure to scram. The CRS directs the P603 operator to perform FSQ 1-
- 8. While performing FSQ 1-8 control rods insert and the following conditions exist:
- APRMs are reading downscale.
- The IRMs have been inserted and are reading between Range 4 and 5 and lowering.
- RPV Pressure is 950 psig and being maintained by Main Turbine Bypass Valves.
- SLC has not been injected.
- 7 Control Rods are FULL OUT, all others are fully inserted.
(1) Is the reactor shutdown under all conditions, and (2) what procedural actions will the CRS direct?
A. (1) NO (2) Inject SLC and insert control rods IAW 29.ESP.03.
B. (1) NO (2) Exit Rx Power leg of 29.100.01A, RPV CONTROL - ATWS, and perform 20.000.21, REACTOR SCRAM.
C. (1) YES (2) Exit ALL legs of 29.100.01A, RPV CONTROL - ATWS, and perform 29.100.01 Sheet 1, RPV CONTROL D. (1) YES (2) Exit Rx Power leg of 29.100.01A, RPV CONTROL - ATWS, and perform 20.000.21, REACTOR SCRAM.
Answer: B 2017 ILO Written Page: 172 of 207 Question 85
Answer Explanation:
Per EPGs, positive confirmation that the reactor will remain shutdown under all conditions without boron is best obtained by determining that no control rod is withdrawn beyond the Maximum Subcritical Banked Withdrawal Position. 7 RODS are out therefore this is not possible. The reactor is below Range 7 on the IRMs and trending down indicating the reactor is shutdown. FSQ override one (1) states that if the reactor is shutdown, exit to the SCRAM AOP.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EPG defination of " remain shutdown under all conditions", it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (RPIS Failure) and then selecting the correct section of EOP to use.
Distracter Explanation:
A. is incorrect because the CRS should exit the ATWS EOP. This answer is plausible because the examinee may determine the correct action is to continue on and inject Boron because the reactor will not remain shutdown under all conditions C. is incorrect because Per EPGs, positive confirmation that the reactor will remain shutdown under all conditions without boron is best obtained by determining that no control rod is withdrawn beyond the Maximum Subcritical Banked Withdrawal Position. 7 rods are full out therefore this is not possible. The examinee may determine the power indications mean the reactor will remain shutdown under all conditions.
D. is incorrect because is incorrect because Per EPGs, positive confirmation that the reactor will remain shutdown under all conditions without boron is best obtained by determining that no control rod is withdrawn beyond the Maximum Subcritical Banked Withdrawal Position. 7 rods are full out therefore this is not possible. The examinee may determine the power indications mean the reactor will remain shutdown under all conditions. The answer is plausible because the examinee may determine that exiting the ATWS EOP and entering the normal RPV Control EOP is the correct course of action.
Plant Procedures 29.100.01 SH 1A 29.100.01 SH 6 BWROG EPG App B NUREG 1123 KA Catalog Rev. 2 295015 AA2. Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM :
295015 AA2.01 Reactor power 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 173 of 207 Question 85
86 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK Earlier in shift the following instruments failed high:
- B21-N090C DIV 1 CS/RHR PERMISSIVE MAIN STM PRESSURE XMTR
- B21-N090D DIV 2 CS/RHR PERMISSIVE MAIN STM PRESSURE XMTR Shortly after identifying the instrument failure, an earthquake occurs. Both Recirc Pumps trip, and Drywell pressure rises causing a reactor scram.
Post reactor scram plant conditions are as follows:
- RPV Level is 164 inches and lowering.
- RPV pressure is 350 psig and lowering rapidly.
- Drywell pressure is 18.3 psig and rising.
- E1150-F017A & B, LPCI Loop Outboard Injection Iso Valves are open Which of the following identifies both (1) the current RHR System valve lineup and (2) the CRS direction to the CRLNO concerning RHR System operation?
A. (1) E1150-F015A & B, LPCI Loop Inboard Isolation Valves, are closed.
(2) Manually open E1150-F015 on the loop desired for injection per 23.205, RHR System.
B. (1) E1150-F015A & B, LPCI Loop Inboard Isolation Valves, are open.
(2) Maintain RPV Water Level 173 to 214 inches using LPCI per EOP 29.100.01, Sheet 1.
C. (1) E1150-F015A & B, LPCI Loop Inboard Isolation Valves, are closed.
(2) Perform Forced LPCI Loop Select Logic Operation on the desired loop per 23.205, RHR System.
D. (1) E1150-F015A & B, LPCI Loop Inboard Isolation Valves, are open.
(2) Manually close E1150-F015 on the loop desired for Containment Cooling per 23.205, RHR System.
Answer: A 2017 ILO Written Page: 174 of 207 Question 86
Answer Explanation:
Reactor Steam Dome Pressure - Low is an input into the LPCI Loop Select Logic. If pressure is <906 psig, LPCI will align for injection. E1150-F015 A/B gets an open signal from LPCI Loop Select, but will not open until reactor pressure is <461 psig (permissive). In this case, the logic will never "see" reactor pressure
<906 psig, so the logic will not initiate. Since E1150-F015A/B will not get an open signal from the logic, it remains closed after the 461 psig permissive is met and must be reopened per 23.205.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of (Need for injection per EOP Sheet 1), it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (Nuclear boiler instrument failures) and then taking action based an SOP to meet then needs of the EOP to mitigate the event.
Distracter Explanation:
B. is incorrect because E1150-F015A & B are closed. If the examinee assumes this is the correct lineup, actions from the EOP are logical.
C. is incorrect because forced LPCI loop select is not required. Both loops are operable, and E1150-F015 can be manually opened. If examinee assumes E1150-F015A & B cannot be opened, forced logic operation would be logical.
D. is incorrect because E1150-F015A & B are closed. If the examinee assumes this is the correct lineup, actions prevent injection from the opposite loop and set up for torus cooling are logical.
Reference Information:
23.601 pg 24 - Reactor Steam Dome Pressure - Low is an input into the LPCI Loop Select Logic. It initiates if reactor pressure is <906 psig and one or both RR Pumps are not running.
ST-OP-315-0041-001 (Fig 18) shows the LPCI Loop Selection Logic - if <2 RR Pumps are not running, and rx pressure is >906 psig, the selection logic will not actuate.
23.205 - lists sections for Manual LPCI Initiation (which requires opening E1150-F015A/B) and Forced LPCI loop Selection Logic operation (plant conditions do not meet required prerequisites).
Plant Procedures 23.205 NUREG 1123 KA Catalog Rev. 2 203000 A2. Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
203000 A2.10 Nuclear boiler instrument failures G2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 175 of 207 Question 86
87 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is operating in MODE 4 when the following alarms occur:
- 1D6, DIV 1 CSS LOGIC POWER FAILURE
- 1D13, TESTABILITY DIV 1 ECCS LOGIC/POWER FAILURE Which one of the following describes the impact of this indication on the Core Spray System ONLY, and what action is required?
A. Automatic operation of Div 1 CSS will be inhibited but manual injection remains functional. Core Spray is required to be OPERABLE in the present MODE. Div 1 CSS must be restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B. Automatic and manual operation of Div 1 CSS will be inhibited. Only one division of Core Spray is required to be OPERABLE in the present MODE. Technical Specification LCO 3.5.2 is satisfied indefinitely.
C. Automatic and manual operation of Div 1 CSS remain functional. Core Spray Loop A is assured to produce the required flow and satisfies the Technical Specification LCO 3.5.2 indefinitely.
D. Automatic and manual operation of Div 1 CSS remain functional but E2150-F031A, DIV 1 CS Pumps Min Flow Valve, Auto closes. Core Spray Loop A is NOT assured to produce the required flow. Div 1 CSS must be restored in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Answer: B 2017 ILO Written Page: 176 of 207 Question 87
Answer Explanation:
The alarms indicate that the DC power to CSS Logic A has had a failure. This will prevent Automatic and Manual injection of Div 1 CSS. The pumps can be manually started but the injection valves are incapable of opening. One division of CSS is required to be operable in Mode 4 (assuming a LPCI subsystem is operable). Tech Spec 3.5.2 is satisfied.
10 CFR 55.43(b)(2) SRO JUSTIFICATION This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of applying T.S. 3.5.2, it is not related to < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action, cannot be answered LCO/TRM information listed above-the-line, and cannot be answered by solely by knowing the TS Safety Limits. The SRO must assessing plant conditions and then apply Required Actions of T.S. and use the requirements of Conditions Distractor Explanation:
A. is incorrect because the injection valves F004A and F005A are interlocked closed and require a signal from CSS Logic A to open.
C. is incorrect because DC Logic power has been lost and both Auto and Manual injection will be inhibited.
D. is incorrect because DC Logic power has been lost and both Auto and Manual injection will be inhibited.
Reference Infromation:
ARP 1D6 Tech Spec 3.5.2 Plant Procedures 01D06 NUREG 1123 KA Catalog Rev. 2 209001 A2. Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
209001 A2.04 D.C. failures Technical Specifications 3.5.2 ECCS Shutdown 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.
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Bank High Selected for Early Review (NRC)
SRO 2017 ILO Written Page: 177 of 207 Question 87
88 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: NEW Due to an error in the calibration procedure for the RPS Reactor Vessel Steam Dome Pressure -
High instruments, the high pressure trip setpoint for all four channels were adjusted such that the channels would not trip until Reactor Vessel Steam Dome Pressure is 1300 psig.
(1) Given the listed conditions will RPS trip prior to reaching the Reactor Coolant System Pressure Safety Limit?
(2) Which of the following will satisfy the requirements of T.S. 3.3.1.1 Reactor Protection System (RPS) Instrumentation Required Action C.1 ONLY?
A. (1) Yes (2) B21-N078A and B21-N078C Set to trip at 1100 psig B. (1) Yes (2) B21-N078A and B21-N078D Set to trip at 1100 psig C. (1) No (2) B21-N078A and B21-N078C Set to trip at 1100 psig D. (1) No (2) B21-N078A and B21-N078D Set to trip at 1100 psig Answer: B 2017 ILO Written Page: 178 of 207 Question 88
Answer Explanation:
To restore trip capability for Reactor Vessel Steam Dome Pressure - High, at least one of the trip systems must be restored (trip setpoint less than 1113 psig psig). To restore a trip system the two channels for that trip system [(A or C) and (B or D)] must be restored.
10 CFR 55.43(b)(2) SRO JUSTIFICATION This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of applying T.S. 3.3.1.1 C.1 Basis. The answer is related to < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action however this knowledege alone cannot answer the question, cannot be answered LCO/TRM information listed above-the-line, and cannot be answered by solely by knowing the TS Safety Limits (RO half required by the K/A). The SRO must assessing plant conditions and then apply Required Actions of T.S. and use the requirements of Conditions to determine the correct alignment of RPS channels to meet the intent of T.S. 3.3.1.1 C.1 Distracter Explanation:
Distracters are plausible and incorrect based on logic listed above and how the T.S. Basis for RPS defines a trip system for RPS required by T.S. 3.3.1.1 C.1 Reference Information:
23.601 Pg 10 (Reactor Vessel Steam Dome Pressure - High)
T.S. 3.3.1.1 COND C (requirement) + BASIS (T.S.B 3.3.1.1)
S.L 2.1.2 NUREG 1123 KA Catalog Rev. 2 212000 RPS G2.2.22 Knowledge of limiting conditions for operations and safety limits Technical Specifications 3.3.1.1 Reactor Protection System (RPS) Instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.
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High New Selected for Early Review (NRC)
SRO 2017 ILO Written Page: 179 of 207 Question 88
89 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The reactor has scrammed due to a coolant leak and a complete loss of the Reactor Feed System. The crew has completed the SCRAM reports and entered the EOPs.
- All Core Spray Pumps are running
- All RHR pumps are running.
- HPCI is injecting.
- RCIC is injecting.
- Drywell pressure is 2.2 psi and increasing at 0.1 psi/minute.
- RPV level is 25 inches and is increasing at 5.0 inches/minute.
- ADS timers are timing down and indicate 93 seconds.
(1) Assuming NO operator action and all trends continue at current rates, what is the status of the ADS SRVs in 2 minutes?
AND (2) Which of the following EOP steps must the CRS direct now?
A. (1) Open (2) Inhibit ADS B. (1) Closed (2) Inhibit ADS C. (1) Open (2) L/U, start pumps, and maximize inj flow with sources from Tables 1 and 3.
D. (1) Closed (2) L/U, start pumps, and maximize inj flow with sources from Tables 1 and 3.
Answer: B 2017 ILO Written Page: 180 of 207 Question 89
Answer Explanation:
In 93 seconds ADS will NOT open the 5 ADS SRV because RPV level will be (25 inches + (93 sec X (1 min / 60 sec) X 5 inches/min) = 32.75 inches, because Level 1 is 31.8 inches the level 1 permissive drops out and the ADS timers reset. See I-2095-06 for DIV 1 explanation, DIV 2 explanation is the same.
Based on conditions EOP Sheet 1 (L-5) is met so ADS must be inhibited.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOP Sheet 1, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (ADS initiation signals present) and then taking action based on the requirements of EOP Sheet 1.
Distracter Explanation:
Distracters are incorrect and plausible based on answer explanation of ADS Valves being open, and the action is the EOP step for recovering from an Emergency Depressurization (Step L-19 sheet 1). The use of Table 3 makes it incorrect.
Reference Information:
I-2095-06 and B21-04 (logic) and 29.100.01 SH 1 Step L-1 to L-6.
Plant Procedures 29.100.01 SH 1 B21-04 AUTOMATIC DEPRESSURIZATION SYSTEM DESIGN BASIS DOC NUREG 1123 KA Catalog Rev. 2 218000 A2. Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
218000 A2.06 ADS initiation signals present 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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High New SRO 2017 ILO Written Page: 181 of 207 Question 89
90 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK During a startup, the North RFPT is operating. The following conditions exist:
- Reactor Power 1% CTP
- Reactor pressure 750 psig
- SULCV 40% open
- SULCV M/A Station AUTO
- North Feedwater Flow Control M/A Station MANUAL The Interruptible Air Supply to the SULCV is now LOST.
(1) How will this failure FIRST affect RPV Water Level, and (2) which ONE of the following actions will CRS direct to mitigate this failure?
A. (1) RPV Water Level will RISE.
(2) PLACE the C32-R620, N21-F403 RPV Startup Level Controller SULCV M/A Station in MANUAL, and lower the OUTPUT signal to CLOSE the SULCV.
B. (1) RPV Water Level will RISE.
(2) TRIP the North RFPT, START the West Standby Feedwater Pump, and control RPV Water Level using N2103-F003, SBFW 4" Disch Flow Ctrl Vlv.
C. (1) RPV Water Level will LOWER.
(2) START the West Standby Feedwater Pump, and control RPV Water Level using N2103-F003, SBFW 4" Disch Flow Ctrl Vlv.
D. (1) RPV Water Level will LOWER.
(2) PLACE the Feedwater Flow Control M/A Station in AUTO, and OPEN N2100-F607, N RFP Disch Line Iso Valve.
Answer: B 2017 ILO Written Page: 182 of 207 Question 90
Answer Explanation:
The SULCV fails OPEN on Loss of Air. With a Reactor Feedwater Pump Turbine operating at a fixed speed (MANUAL) - RPV Water Level will RISE. TRIPPING the RFPT will stop flow through the SULCV and SBFW can be operated to control RPV Water Level.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP 20.129.01 and SOP 23.107, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (Air loss to SULCV) and then taking action based on the requirements of AOP 20.129.01 and SOP 23.107.
Distractor Explanation:
A. Is plausible and incorrect; would be true for auto control signal failure to SULCV. Irrespective of controller output, with Loss of Air, the SULCV will continue to OPEN and RPV Water Level will RISE.
This will NOT mitigate the failure.
C Is plausible and incorrect; but RPV Water Level will RISE.
D. Is plausible and incorrect; but RPV Water Level will RISE.
Reference Information:
AOP 20.129.01 Air loss impact (Enclosures)
Plant Procedures 20.107.01 20.129.01 23.107 NUREG 1123 KA Catalog Rev. 2 259002 A2. Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
259002 A2.05 Loss of applicable plant air systems 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 183 of 207 Question 90
91 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK After a reactor scram from a turbine trip, the following conditions exist:
- Reactor Power is 3%.
- Torus Temperature is 140°F and rising.
- Drywell Pressure is 1.4 psig and steady.
- SLC is injecting.
- Division 2 RHR is in Torus Cooling and Torus Spray.
- RPV water level is currently 63 inches with a directed level band of 50 to 100 inches.
- 29.ESP.11 has been directed, but is NOT complete.
The CRLNO subsequently reports he is out of his water level band LOW. RPV level currently indicates 30 inches.
(1) What is the specific effect of these conditions on DIV 2 RHR, and (2) what actions shall the CRS direct to the CRLNO?
A. (1) E1150-F027B, Div 2 RHR Torus Spray Iso, E1150-F028B, Div 2 RHR Torus Iso Vlv. and E1150-F024B, Div 2 RHR Torus Clg Iso, close.
(2) Terminate and Prevent injection due to exceeding the BIIT.
B. (1) Only E1150-F028B, Div 2 RHR Torus Iso Vlv. closes.
(2) Terminate and Prevent injection due to exceeding the BIIT. Then place DIV 1 RHR in Torus Cooling and Torus Spray.
C. (1) E1150-F027B, Div 2 RHR Torus Spray Iso, E1150-F028B, Div 2 RHR Torus Iso Vlv. and E1150-F024B, Div 2 RHR Torus Clg Iso, close.
(2) Change water level band to 0-50 inches. Then place Containment Spray Mode Select Switch in MANUAL, and re-open F027B, F028B, and F024B.
D. (1) Only E1150-F027B, Div 2 RHR Torus Spray Iso, and E1150-F024B, Div 2 RHR Torus Clg Iso, close.
(2) Change water level band to 0-50 inches. Then place Containment Spray 2/3 Core Height Override Switch in MANUAL OVERRIDE, and re-open F027B and F024B.
Answer: C 2017 ILO Written Page: 184 of 207 Question 91
Answer Explanation:
Due to level falling below Level 1, E1150-F027 B, Div 2 RHR Torus Spray Iso, E1150-F028B, Div 2 RHR Torus Iso Vlv, and E1150-F024 B, Div 2 RHR Torus Clg Iso, close. To restore, these valves must be re-opened after Containment Spray Mode Select Switch is placed in MANUAL. Due to 29.ESP.11 not being completed, MSIVs will go closed. ODE-10 states expected level band for this condition is 0-50 inches (Reactor Power >3% & heat discharging to Torus: 0-50 inches).
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOP SH 1 & SH 2, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (RHR/LPCI: TORUS/SUPPRESSION POOL SPRAY MODE) and then taking action based on the requirements of EOP SH 1 & SH 2 and 23.205.
Distractor Explanation:
A. RHR isolation valves will close because of L1 but Torus Temperature is not greater than the BIIT.
B All RHR isolation valves will close. Torus temp is below the BIIT.
D. Containment Spray 2/3 Core Height Override Switch is not placed in MANUAL OVERRIDE until RPV level is below Level 0.
Reference Information:
23.205, Enclosure A, 29.100.01 SH1A Plant Procedures 23.205 29.100.01 SH 1A NUREG 1123 KA Catalog Rev. 2 203000 A2. Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
203000 A2.03 Valve closures 230000 A2. Ability to (a) predict the impacts of the following on the RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
230000 A2.03 Valve closures 295037 EK2. Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following:
295037 EK2.09 Reactor water level 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 185 of 207 Question 91
92 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: BANK Which one of the following completes the statement which describes the bases of the Limiting Condition for Operation which requires the Main Turbine Bypass valves and the Moisture Separator Reheater to be operable above 25% power?
An operable Main Turbine Bypass system and the Moisture Separator Reheater will ensure A. the plant cool down limits are not exceeded during all turbine trip events.
B. RPV pressure scram setpoints are not exceeded during all turbine trip events.
C. the MCPR safety limit is not exceeded during rapid over pressurization transients.
D. the APLHGR safety limit is not exceeded during rapid over pressurization transients.
Answer: C 2017 ILO Written Page: 186 of 207 Question 92
Answer Explanation:
The tech spec bases describes the systems as being required to maintain MCPR within the safety limit for over pressurization transients.
Closed Reference 10 CFR 55.43(b)(2) SRO JUSTIFICATION This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of T.S.B, it is not related to < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action, cannot be answered LCO/TRM information listed above-the-line, and cannot be answered by solely by knowing the TS Safety Limits. The SRO must have knowledge of the Basis for T.S.
3.7.6.
Distracter Explanation:
A is plausible, the bypass valves are used to cooldown the plant after a scram but the system is not designed to prevent exceeding cool down limits.
B is plausible; The bypass and MSR system will limit peak reactor pressure during turbine trip events but will not prevent reactor pressure from rising above the scram setpoint for ALL turbine trip events.
D is plausible; APLHGR is a thermal limit, but not a safety limit. It is required to be maintained above 25%
power but not for over pressure transients.
Reference Information:
T.S.B 3.7.6 NUREG 1123 KA Catalog Rev. 2 245000 Main Turbine Generator and Auxiliary System G2.2.25 Knowledge of bases in Technical Specifications for limiting conditions for operations and safety limits.
Technical Specifications 3.7.6 The Main Turbine Bypass System and Moisture Separator Reheater (Basis) 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (3) Mechanical components and design features of reactor primary system.
10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.
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Bank High SRO 2017 ILO Written Page: 187 of 207 Question 92
93 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW The plant is operating at 85% power.
At 1020 a seismic event occurs.
At 1030 IPCS ARM display indicates:
2017 ILO Written Page: 188 of 207 Question 93
At 1040 IPCS ARM display indicates:
2017 ILO Written Page: 189 of 207 Question 93
Based on the given indications and assuming all trends continue with no operator action:
At time 1050 (1) what is the impact on Secondary Containment and (2) which of the following lists ALL REQUIRED actions for Secondary Containment Control to be directed by the CRS?
A. (1) 2 areas will exceeded MAX NORM.
(2) Isolate ALL systems discharging into this area except systems required by EOPs or damage control ONLY.
B. (1) 1 area will exceeded MAX SAFE.
(2) Isolate ALL systems discharging into this area except systems required by EOPs or damage control ONLY.
C. (1) 2 areas will exceeded MAX NORM.
(2) Isolate ALL systems discharging into this area except systems required by EOPs or damage control.
AND PERFORM 29.100.01 SH 1 entering at step RC-1.
D. (1) 1 area will exceeded MAX SAFE.
(2) Isolate ALL systems discharging into this area except systems required by EOPs or damage control.
AND PERFORM 29.100.01 SH 1 entering at step RC-1.
Answer: D 2017 ILO Written Page: 190 of 207 Question 93
Answer Explanation:
Based on trends:
CSS Div 2 is increased 19.9°F so temperature will be 145.9°F which is below max norm (148°F)
HPCI PMP is increased by 24.8°F so temperature will be 212°F which is above max safe (210°F)
Steam Tunnel is increased 5.0°F so temperature will be 142.5°F which is below max norm (160°F)
Based on this ONE area is above MAX SAFE and with increase in HPCI Room rad to 65.65 (indication of primary system discharging to secondary containment)
EOP flow chart require steps SCT-1 through SC-6.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOP Secondary Containment Control, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (Max Safe or Max Norm) and then taking action based on the requirements of EOP Secondary Containment Control.
Distactor Explanation:
Distractor are incorrect and plausable based on knowing values for max safe/norm and required actions of EOP.
Reference Information:
29.100.01 SH 5 Plant Procedures 29.100.01 SH 5 NUREG 1123 KA Catalog Rev. 2 290001 A2. Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
290001 A2.05 High area temperature.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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SRO 2017 ILO Written Page: 191 of 207 Question 93
94 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK Which of the following job functions is the responsibility of the Refuel Floor Supervisor during a refueling outage?
A. Maintain control of personnel entering the Refuel Floor.
B. Provide the lead responsibility for all fuel movement instructions and plans.
C. Determine if an evacuation is neccessary whenever a CAM or ARM alarms.
D. Act as Reactivity Management SRO during Control Rod testing in a cell that contains fuel.
Answer: D 2017 ILO Written Page: 192 of 207 Question 94
Answer Explanation:
Refuel Floor Supervisor is an SRO responsibility. MOP13 states that, during a refueling outage, the Refuel Floor Supervisor shall be on the Refuel Floor during fuel movements, and in the Control Room (as Reactivity Management SRO) during control rod testing in a cell that contains fuel.
10 CFR 55.43(b)(7) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because this question is about a refuel floor SRO responsiblity that is NOT a share responsibilty with an RO.
Distracter Explanation:
All distractors are plausible because they are all functions/responsibilities performed during refueling outages.
A is the responsibility of the Refuel Floor Coordinator.
B. is a function of the Supervisor, Reactor Engineering.
C. is the responsibility of Radiation Protection Reference Information:
MOP 13 Section 3.2.1 Question Cognitive Level Knowledge Plant Procedures MOP13 Question Use ILO NUREG 1123 KA Catalog Rev. 2 G2.1.35 Knowledge of the fuel-handling responsibilities of SROs 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (7) Fuel handling facilities and procedures.
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Bank Low SRO 2017 ILO Written Page: 193 of 207 Question 94
95 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: BANK Night Orders are issued by the __(1)__ /delegate for a specified time which will not normally exceed __(2)__.
A. (1) Shift Manager (2) 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> B. (1) Manager-Nuclear Operations (2) 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> C. (1) Shift Manager (2) 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> D. (1) Manager-Nuclear Operations (2) 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> Answer: D 2017 ILO Written Page: 194 of 207 Question 95
Answer Explanation:
Per MOP01 3.20.1 The Manager Nuclear Operations/delegate shall issue Night Orders to be in effect for the period of time specified in the document. This time may be up to, but should not normally exceed, 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of MOP01 it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on knowledge of administrative procedures that specify normal procedures.
Distracter Explanation:
Shift Manager is a plausible distractor, and a time limit of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is reasonable.
Reference Information:
MOP01, section 3.20.1 Plant Procedures MOP03 - Conduct Of Operations NUREG 1123 KA Catalog Rev. 2 G2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, operations memos, etc.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank Low SRO 2017 ILO Written Page: 195 of 207 Question 95
96 Points: 1.00 Difficulty: 2.00 Level of Knowledge: High Source: BANK It is March, and the plant is operating at 100% power. The CSS calls to report weather conditions will result in sustained winds of 43-48 mph and gusts > 60 mph.
In accordance with PSA and Maintenance Rule On-Line Risk Management Guidelines, the PSA Model Event should be manually adjusted to account for the changing system conditions.
The appropriate condition is ___(1)___ , and the EOOS Adjustment Factor should be increased to ___(2)___ .
A. (1) High Wind Advisory (2) Elevated B. (1) High Wind Advisory (2) High C. (1) High Wind Watch/Warning (2) Elevated D. (1) High Wind Watch/Warning (2) High Answer: D 2017 ILO Written Page: 196 of 207 Question 96
Answer Explanation:
Per MMR APP H Section C In the event that conditions arise that have or may have an adverse impact on an initiating event, it is prudent to manipulate the PSA model to account for these conditions. This requires manual adjustment of the frequency of one or more of the initiating events assumed to occur in the PSA model.
Based on this the SRO would need to evaluate wind conditions in the table listed in Section C and find the following results:
For Condition:
High Wind Watch/Warning (sustained winds > 40 mph, wind gusts > 58 mph).
Initiating Event Impacted:
Loss of Offsite Power EOOS Frequency Adjustment Increase to High 10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of actions required by a conduct manual, it is not related to immediate actions and the entry conditions are not relevant or leading to answer. The answer to this questions is based on assessing plant conditions (wind speed) and then selecting a procedure section to with which to proceed.
Distractor Explanation:
Distracter combinations are incorrect and plausible based on interpretation the table in MMR APP H Section C.
Reference Information:
MMR APP H Section C Plant Procedures MMR Appendix H Online Core Damage Risk Management Guidelines NUREG 1123 KA Catalog Rev. 2 G2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 197 of 207 Question 96
97 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK The plant is running at rated power with the HPCI system inoperable.
While performing 24.137.01, OPERABILITY OF 480V SWING BUS 72CF AUTOMATIC THROWOVER SCHEME, the throwover failed to occur when the Auto Throwover Permissive Test switch at Bus 72C, Position 1A was taken to CLOSE. The SRO has entered Tech Spec 3.0.3.
Maintenance personnel have submitted Form MMA26001, TROUBLESHOOTING DATASHEET, for approval with the following information:
- The troubleshooting boundary is Bus 72C, Position 1A cubicle plane
- A flashlight must break the cubicle plane to perform an adequate visual inspection based on orientation Based on the information given, what is the LOWEST level of approval required?
A. Control Room NSO B. Control Room Supervisor C. Shift Manager D. Maintenance Manager Answer: B 2017 ILO Written Page: 198 of 207 Question 97
Answer Explanation:
MMA26 Enclosure C designates this activity to be Category C based on it being High Risk and the Rigor Matrix. The Rigor Matrix allows Category C based on Visual Inspection with barrier intrusion. Category C can be approved by an SRO.
Distractor Explanation:
A. The Control Room LNO CAN NOT approve the Troubleshooting data sheet. They are only required to be RO licensed.
C/D. The Maintenance Manager and SM CAN approve the Troubleshooting data sheet, but they are not the LOWEST level. MM approval is required for Categories A/B.
Reference Information:
MMA 26 Encl C Plant Procedures MMA26 - Troubleshooting TECH SPECS NUREG 1123 KA Catalog Rev. 2 G2.2.20 Knowledge of the process for managing troubleshooting activities.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 199 of 207 Question 97
98 Points: 1.00 Difficulty: 2.00 Level of Knowledge: Low Source: NEW The Emergency Director will make a Protective Action Recommendation (PAR) decision based on an estimate of radiation exposure of __(1)__ or __(2)__ that an individual member of the PUBLIC might receive.
A. (1) TEDE 1 rem (2) Adult Thyroid 5 rem B. (1) TEDE 1 rem (2) Adult Thyroid 10 rem C. (1) TEDE 5 rem (2) Adult Thyroid 5 rem D. (1) TEDE 5 rem (2) Adult Thyroid 10 rem Answer: A 2017 ILO Written Page: 200 of 207 Question 98
Answer Explanation:
Per EP-545:
PAR decision is based on an estimate of radiation exposure an individual might receive over a projected period in comparison to a Protective Action Guideline (PAG).
PAG values are as follows:
TEDE 1 rem Adult Thyroid 5 rem 10 CFR 55.43(b)(4) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because this question is about radiation values used for a SRO responsibility to determine a PAR that is NOT a share responsibility with an RO or are values that are required to be known by the RO for radiological safety stay times.
Distracter Explanation:
Distracter are incorrect and plausible because values are from other radiological limits Reference Information:
EP-545 Section 4.3.2 ILO 2017 Exam Notes:
Question is about limits in relation to an emergency and public dose. While NGET knowlegde of dose limits are known to rad workers, rad workers do not know PARs and how dose is related to PARs. Dose numbers for distractors are common numbers used in relation to dose limits, to make them plausible.
Plant Procedures EP-545 NUREG 1123 KA Catalog Rev. 2 G2.3.4 Knowledge of radiation exposure limits under normal and emergency conditions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
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Low New SRO 2017 ILO Written Page: 201 of 207 Question 98
99 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: BANK A fire has started in the H11-P603 panel in the control room. The fire could not be extinguished by control room personnel, and personnel evacuation has been directed.
The manual scram push buttons have been depressed; however, the switches malfunctioned, and the reactor did not scram.
Which one of the following describes the alternate actions which shall be directed by the CRS to shut down the reactor for the given conditions during the fire?
A. Place the Mode Switch in Shutdown.
B. Turn Off circuits 5 & 6 from Dist Cab 2PA-2 and 2PB-2 C. De-energize both Power Range Neutron Monitoring systems from the RPS power panel in the DC MCC area.
D. Fully de-energize all RPS system loads from the RPS power panel in the DC MCC area.
Answer: C 2017 ILO Written Page: 202 of 207 Question 99
Answer Explanation:
A fire in the control room (Zone 12) would require entry into procedure 20.000.18, CONTROL OF THE PLANT FROM THE DEDICATED SHUTDOWN PANEL as directed from procedure 20.000.22, PLANT FIRES. The first action to scram the reactor using the manual scram push buttons is the preferred method in accordance with 20.000.18. De-energizing both power range neutron monitoring systems from the RPS power panel in the DCMCC area is the alternate method directed.
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP 20.000.18, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (plant damage) and then the correct mitigating action to take.
Distractor Explanation:
A. is incorrect but plausible because placing the mode switch in shutdown would be a reasonable alternate method which would scram the reactor. For the conditions given in the stem, this action would not be appropriate and not directed by the controlling procedure 20.000.18.
B. is incorrect but plausible because this action is directed to remove power to the SRVs to preclude an uncontrolled RPV blowdown due to spurious actuation of SRVs (Prompt Opertor Action).
D. is incorrect but plausible because this action is directed as another alternate method of reactor shutdown for control room evacuation, but not as a result of a fire. This action is directed by 20.000.19, CONTROL OF THE PLANT FROM THE REMOTE SHUTDOWN PANEL. This action would also cause containment isolation, complicating the actions required by 20.000.18.
Reference Information:
20.000.18 Condition B.1 Plant Procedures 20.000.18 20.000.19 20.000.22 NUREG 1123 KA Catalog Rev. 2 G2.4.25 Knowledge of fire protection procedures.
10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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Bank High SRO 2017 ILO Written Page: 203 of 207 Question 99
100 Points: 1.00 Difficulty: 3.00 Level of Knowledge: High Source: NEW Following a transient in which RPV level dropped below the top of active fuel (TAF), plant conditions are as follows:
- RPV level is being controlled at 197 inches.
- ALL Control Rods are inserted.
- Containment Venting is being performed to maintain Torus Pressure below the Primary Containment Pressure Limit.
- Containment High Range Radiation Monitors indicate 2000 R/hr, 20 minutes after the scram.
In accordance with EP-101, which ONE of the following Emergency Action Levels is appropriate based solely on Fission Product Barrier Degradation?
A. None.
B. Alert FA1.1 - Any Loss OR Any Potential Loss of EITHER Fuel Clad OR Reactor Coolant System.
C. Site Area Emergency FS1.1 - Loss OR Potential Loss of Any Two Barriers.
D. General Emergency FG1.1 - Loss of Any Two Barriers AND Potential Loss of Third Barrier Answer: C 2017 ILO Written Page: 204 of 207 Question 100
Answer Explanation:
Per EP-101, Encl B 2017 ILO Written Page: 205 of 207 Question 100
10 CFR 55.43(b)(5) SRO Justification:
This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of Emergency Classifaction from EP-101, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (plant damage) and then selecting the correct section of EP-101 to use.
Distractor Explanation:
Distractors are incorrect and plausible based on applying EP-101,Encl B Reference Information:
EP-101, Encl B Plant Procedures EP-101 NUREG 1123 KA Catalog Rev. 2 G2.4.41 Knowledge of the emergency action level thresholds and classifications 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
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High New SRO 2017 ILO Written Page: 206 of 207 Question 100
2017 ILO Written Page: 207 of 207 Question 100